underground disposal of radioactive wastes
TRANSCRIPT
Underground Disposal of Radioactive Wastes
Vol. I^ у PROCEEDINGS OF A SYMPOSIUM, OTANIEMI, 2-6 JULY 1979
JOINTLY ORGANIZED BY IAEA AND NEA (OECD)
* * * * 4 *
Deep geological ^4
Earthen material
Low-level-wastecontainers
j Host rock % (w ith fluids)
Backfilled tunnel
Conditioned alpha-bearing wastes
Conditioned high-level wastes
c——-5 C j_Atmosphere
Land surface Surface waters
Aquifers
Sedimentary layers
(highly variable)
M. A. Fpradwfr
INTERNATIONAL ATO M IC ENERGY AG ENCY, V IEN NA, 1980
UNDERGROUND DISPOSAL
OF
RADIOACTIVE WASTES
VOL. I
PROCEEDINGS SERIES
UNDERGROUND DISPOSAL OF
RADIOACTIVE WASTESPROCEEDINGS OF A SYMPOSIUM ON
THE UNDERGROUND DISPOSAL OF
RADIOACTIVE WASTES
JOINTLY ORGANIZED BY THE
INTERNATIONAL ATOMIC ENERGY AGENCY
AND THE OECD NUCLEAR ENERGY AGENCY
AND HELD AT
OTANIEMI, FINLAND, 2 - 6 JULY 1979
In two volumes
VOL.I
INTERNATIONAL ATOMIC ENERGY AGENCY
VIENNA, 1980
UNDERGROUND DISPOSAL OF RADIOACTIVE WASTES, VOL. I
IAEA, VIENNA, 1980
STI/PUB/528
ISBN 92-0-020180-2
© IA E A , 1980
Perm ission to rep ioduce o r translate the in fo rm a tio n co n ta in ed in this pu b lic a t io n m ay be o b ta in ed by
w riting to the In te rn a t io n a l A to m ic Energy Agency , Wagramerstrasse 5, P .O . B ox 100, A-1400 V ie nna , A ustr ia .
P rin ted by the IA E A in A ustria
J u ly 1980
FOREWORD
Disposal of radioactive waste is an issue ¡of central interest for the accept
ance and further industrial development of nuclear power. With today’s
technology, the most feasible option for the safe disposal of these wastes is to
deposit them underground in an appropriately conditioned form at suitable
sites.
Disposal of low- and intermediate-level radioactive wastes by shallow land
burial, emplacement in suitable abandoned mines, or by deep well injection and
hydraulic fracturing, has been practised in various countries for many years. In
recent years considerable efforts have been devoted in most countries that have
nuclear power programmes to developing and evaluating appropriate disposal
systems for radioactive wastes, in particular for high-level ánd transuranium-
bearing wastes, and to studying the potential for establishing repositories in
geological formations underlying their national territories.
In view of this the IAEA felt it was timely to hold a symposium to collect
new information and review current developments in this field. The symposium
was organized jointly by the IAEA and the OECD Nuclear Energy Agency in
co-operation with the Geological Survey of Finland, at the Technical University
of Helsinki, in Otaniemi, Finland. It was attended by about 400 participants
from 32 countries and four international organizations. A total of 68 papers
was presented in ten sessions covering the following topics: national pro
grammes and general studies; disposal of solid waste at shallow depth and in
rock caverns; disposal of liquid waste by deep well injection and hydraulic
fracturing; disposal in salt formations, crystalline rocks and argillaceous sedi
ments; thermal aspects of disposal in deep geological formations; radionuclide
migration studies; and safety assessment and regulatory aspects. While the
disposal of high-level and alpha-bearing wastes arising from the management of
spent nuclear fuel was the central subject of the symposium, many papers also
dealt with matters concerning the disposal of low- and intermediate-level
wastes.
The papers and discussions published in the present Proceedings provide
an authoritative account of the status of underground disposal programmes
throughout the world in 1979. They evidence the experience that has been
gained and the comprehensive investigations that have been performed to study
various alternative possibilities for the underground disposal of radioactive
waste since the last IAEA/NEA symposium on this topic (Disposal of Radio
active Waste into the Ground) was held in Vienna in 1967. The symposium
showed an impressive variety of viable disposal options. It indicated also the
trend to develop a broad scientific base behind the concept of geological waste
disposal. Different approaches are being investigated for the emplacement of
the various waste forms in various rock types. Many geological environments
exist with the capability of providing safe isolation for all types of radioactive
waste.
It is hoped that these Proceedings, together with other documents published
within the Agency’s Underground Disposal Programme, will assist and guide
further national and international efforts in this important field.
EDITORIAL NOTE
The papers and discussions have been edited by the editorial staff o f the International
Atomic Energy Agency to the extent considered necessary for the reader’s assistance. The views
expressed and the general style adopted remain, however, the responsibility o f the named authors
or participants. In addition, the views are not necessarily those o f the governments of the
nominating Member States or of the nominating organizations.
Where papers have been incorporated into these Proceedings without resetting by the Agency,
this has been done with the knowledge of the authors and their government authorities, and their
cooperation is gratefully acknowledged. The Proceedings have been printed by composition
typing and photo-offset lithography. Within the limitations imposed by this method, every effort
has been made to maintain a high editorial standard, in particular to achieve, wherever practicable,
consistency of units and symbols and conformity to the standards recommended by competent
international bodies.
The use in these Proceedings o f particular designations o f countries or territories does not
imply any judgement by the publisher, the IAEA, as to the legal status o f such countries or
territories, o f their authorities and institutions or of the delimitation of their boundaries.
The mention of specific companies or of their products or brand names does not imply any
endorsement or recommendation on the part of the IAEA.
Authors are themselves responsible for obtaining the necessary permission to reproduce
copyright material from other sources.
CONTENTS OF VOLUME I
PROGRAMMES AND GENERAL STUDIES (Sessions I and И)
Warranty obligations for the management and underground disposal
of radioactive waste (IAEA-SM-243/119) ................................................ 3
P. Jauho, P. Silvennoinen
Discussion .................................................................................................. 18
The USA’s programme for disposal of radioactive wastes
(IAEA-SM-243/77) ..................................................................................... 19
J.M. Batch, C.A. Heath
Discussion .................................................................................................. 29
Программа исследовательских работ по окончательному захоронению высокоактивных отходов в глубокие слабопроницаемые геологические
формации (IAEA-SM-243/115) .................................................................... 31
О. J1.Кедровский, Е.А.Леонов, Н.М.Ромадин, И.Ю.Шишиц
(Research programme on the disposal o f high-level wastes in deep
geological formations with low permeability)
Discussion .................................................................................................. 39
National policy for underground disposal o f radioactive wastes in theUnited Kingdom (IAEA-SM-243/30) ........................................................ 41
F.S. Feat es
Discussion .................................................................................................. 47
Underground disposal of radioactive wastes in India: past experience
and future planning (IAEA-SM-243/158) ................................................ 49
K.T. Thomas, N.S. Sunder Rajan, К. Balu, M.P.S. Ramani
Discussion .................................................................................................. 63
The Federal Republic of Germany’s programme for the disposal of
radioactive waste (IAEA-SM-243/95) ........................................................ 65
K. Kühn, R.P. Randl, H. Rôthemeyer
Discussion .................................................................................................. 76
Development of deep underground disposal for Canadian nuclear
fuel wastes (IAEA-SM-243/167) ................................................................ 79
S.R. Hatcher, S.A. Mayman, M. Tomlinson
• Discussion .................................................................................................. 90
Concept and realization programme for final storage o f nuclear wastein Switzerland (IAEA-SM-243/160)............................................................ 93
H. Issler, R.H. Beck, H. Zünd
Discussion ................................................................................................. 103
A strategy for the disposal of radioactive wastes in Italy
(IAEA-SM-243/67) ..................................................................................... 105
M. Mittempergher
Les activités de recherche et de développement des communautés
européennes en matière d’évacuation des déchets radioactifs dans
les formations géologiques (IAEA-SM-243/128) ....................................... 115
P. Venet, E. Delia Loggia, W. Falke, B. Haijtink, Ph. Masure
Invited paper:
Submarine geologic disposal of nuclear waste (IAEA-SM-243/99) ................ 131
CD. Hollister, B.H. Corliss, D R. Anderson
Поверхностные хранилища остеклованных высокоактивных отходов
(IAEA-SM-243/112) .................................................................... ................ 141
A.Н.Кондратьев, В.В.Куличенко, И.И.Крюков, Н.В.Крылова,
B. И. Парамошкин, М. В. Страхов
(Near-surface storage facilities for vitrified high-level wastes)
UNDERGROUND DISPOSAL OF LIQUID WASTE, DISPOSAL Oh SOLID
WASTE AT SHALLOW DEPTH AND IN ROCK CAVERNS (Session HI)
Принципы оценки надежности подземного захоронения радиоактивныхжидких отходов в глубокие геологические формации и пути ее повышения(IAEA-SM-243/110) ....................................................................................... 153
О.ЛКедровский, М.К.Пименов, Н.А.Раков, А.И.Рыбалъченко,
Ф. П. Юдин
(Principles for evaluating the reliability o f underground disposal of
liquid radioactive wastes in deep geological formations and ways of
improving reliability)
Discussion ................................................................................................. 168
Waste disposal by shale fracturing at Oak Ridge National Laboratory
(IAEA-SM-243/42) ..................................................................................... 171
H.O. Weeren
Discussion ................................................................................................. 176
Физико-химические процессы при удалении жидких радиоактивных
отходов в глубокие пласты-коллекторы (IAEA-SM-243/113)........... ............... 179
В. И. Спицин, В.Д.Балукова
(Physical and chemical processes occurring as a result o f the disposal
o f liquid radioactive wastes in deep formations)
Discussion ................................................................................................. 190
ALMA — a study of a repository for low- and medium-level waste
in a rock cavern (IAEA-SM-243/66) .......................................................... 193
N. Rydell, 0. Degerman, C. Thegerstrôm, R. Gelin,
M. Cederstrôm
Discussion .....................................................................................................207
Disposal of low- and intermediate-level waste in Czechoslovakia
(IAEA-SM-243/156) ................................................................................... 209
Z. Dlouhy, J. Kortus, E. Malásek, J. Marek, M. Seliga
Discussion .................................................................................................... 218
Shallow land burial: experience and developments at Los Alamos
(IAEA-SM-243/150) ..................................................!..............................221
J.L. Warren
Discussion .....................................................................................................239
Application de la technique des barrières capillaires aux stockages
entranchées (IAEA-SM-243/71 ) .................................................................. 241
D. Rançon
Discussion .....................................................................................................265
DISPOSAL IN DEEP GEOLOGICAL FORMATIONS: EVAPORITES
(Session IV)
Characterization of a site in bedded salt for the isolation of radioactive
wastes (IAEA-SM-243/3 8) ......................................................................... 269
L.R. Hill
Discussion ..................................................................................................286
Site selection, site investigations and design activities in the USA for
nuclear waste repositories in bedded and salt dome formations
(IAEA-SM-243/151) .......................................................... ...................... 289
M. Kehnemuyi, S.C. Matthews
Discussion ..................................................................................................295
Site investigations and conceptual design for the repository in the
nuclear ‘Entsorgungszentrum’ of the Federal Republic of Germany
(IAEA-SM-243/48) ..................................................................................... 297
H. Rôthemeyer
Discussion ..................................................................................................308
Studies on the optimal disposal of radioactive wastes with special
attention to the thermal influence on the surrounding salt bed and to
economic aspects (IAEA-SM-243/162) .................................................... 311
A. S. Kunstman, K.M. Urbanczyk, J.K. Wierzchoñ, J. -taszkiewicz
Discussion ..................................................................................................323
Исследование радиационной стойкости природной каменной соли(IAEA-SM-243/109) ...................................................................................... 325
В .И . Спицын, Л И .Б а р с о в а , С .А .К абак чи , И .И . З я зю л я , И .Е .Л е б е д е в а
(Investigation o f the radiation stability o f natural rock salt)
Summary of United States Geological Survey investigations of fluid-rock-
waste reactions in evaporite environments under repository conditions
(IAEA-SM-243/97) ...................................................................................... 335
D.B. Stewart, B.F. Jones, E. Roedder, R.W. PotterII
Discussion ................................................................................................... 343
Hydrogeological research at the site of the Asse salt mine
(IAEA-SM-243/13) ...................................................................................... 345
H. Batsche, W. Rauert, K. Klarr
Discussion ................................................................................................... 369
Disposal and fixation of medium- and low-level liquid wastes in salt
caverns: in situ solidification (IAEA-SM-243/16) .................................... 371
R. Kôster, R. Kraemer, R. Kroebel
Discussion ........................................... ....................................................... 383
DISPOSAL IN DEEP GEOLOGICAL FORMATIONS:
CRYSTALLINE ROCKS (Session V)
Premières évaluations des possibilités d ’évacuation des déchets radioactifsdans les roches cristallines (IAEA-SM-243/68) ...................................... 387
A. Barbreau, Y. Sousselier, M. Bonnet, J. Margat, P. Peaudecerf
P. Goblet, E. Ledoux, G. deMarsily
Discussion ................................................................................................. 410
Canadian Geoscience research and design concepts for disposal of high-
level waste in igneous rocks (IAEA-SM-243/168) .................................... 413
J.S. Scott, R.G. Charlwood
Discussion ................................................................................................... 439
The Swedish geological programme for the disposal of high-level waste
(IAEA-SM-243/163) ................................................................................. .443
U. Thoregren, K. Ahlbom, G. Gidlund, C.E. Klockars,
K.A. Magnusson, S. Scherman
Discussion ................................................................................................... 453
Review of geological criteria and site selection for high-level
radioactive waste repositories in the United Kingdom
(IAEA-SM-243/29) ..................................................................................... 455
J.D. Mather, D.A. Gray, P.B. Greenwood
Discussion ................................................................................................... 464
Geological disposal of high-level radioactive waste: conceptual
repository design in hard rock (IAEA-SM-243/93) .....................................467
H. Beale, J.R. Griffin, J.W. Davies, W.R. Burton
Discussion ................................................................................................... 477
The nuclear waste disposal study project of the Geological Survey of
Finland (IAEA-SM-243/118) ..................................................................... 479
H. Niini
Bentonite-based buffer substances for isolating radioactive waste products
at great depths in rock (IAEA-SM-243/22)..................................................487
R. Pusch, A. Jacobsson, A. Bergstrom
Discussion ..................................................................................................501
Corrosion studies on copper and titanium-lead canisters for nuclear
waste disposal (IAEA-SM-243/166)............................................................ 503
L.B. Ekbom, K. Hannerz, K.S. Henrikson
Discussion ..................................................................................................514
Chairmen of Sessions and Secretariat of the Symposium.................................517
PROGRAMMES AND GENERAL STUDIES
(Sessions I and II)
Chairmen
Session I
A. BARBREAU
France
Session II
I.S. ZHELUDEV
International Atomic Energy Agency
IAEA-SM-243/119
WARRANTY OBLIGATIONS FOR THE
MANAGEMENT AND UNDERGROUND
DISPOSAL OF RADIOACTIVE WASTE
P. JAUHO, P. SILVENNOINEN
Technical Research Centre of Finland,
Espoo, Finland
Abstract
WARRANTY OBLIGATIONS FOR THE MANAGEMENT AND UNDERGROUND DISPOSAL
OF RADIOACTIVE WASTE.
The need for financial assurances and institutional arrangements for waste management
and disposal is discussed from the viewpoint of public interest. The basic principles stated in the
paper include the requirement of accumulating funds for future contingencies during the active
lifetime of the reactors and the fuel cycle facilities. A governmental role is seen as indispensable
in assuming responsibility over at least the surveillance of underground repositories. The stage
at which the operational responsibility is transferred from the plant operator to the government
is determined in general by the status of the waste conditioning and disposal technology. A brief
survey is presented of the current situation and technical issues. The need for special funds is
discussed as well. For the part of waste management and disposal that will be taken over by the
government an escrow fund should be established. Parallel to this public fund the plant operator
would be obliged to reserve funds and provide guarantees within the company to cover
liabilities for the remaining part of waste management and disposal obligations. A case study is
presented in the paper covering the estimation of the escrow charges for spent fuel or high-level
waste.
1. INTRODUCTION
Due to the long-lasting potential risk involved, waste management has both
philosophical and technical aspects which may be more perplexing than conventional
safety issues. Although the conceivable hazard is not necessarily more threatening
than that encountered with stable toxic waste materials generated in process
industry, nuclear energy is unlikely to be acceptable to society unless adequate
assurances are provided for the future custody of the irradiated fuel and the waste.
The issuance of operating licences for nuclear power plants already implies
that the licensing authority is convinced of the future development of an acceptable
waste disposal technology. Due to the lack of large-scale waste disposal
demonstrations the authority is bound to stipulate provisions in the permit in order
to guarantee the existence of the necessary financial and institutional means at the
time when the disposal is to take place or in case unexpected developments call for
precautions in the storage and disposal schemes.
3
4 JAUHO and SILVENNOINEN
This paper focuses on some of the alternatives for providing the guarantee
arrangements that seem to be necessary in order to prevent society today and
the current generation from passing on excessive risks from nuclear wastes to future
generations. The financial liability is to be borne by the consumers while the longer-
. term risk is to be borne by society through institutional mechanisms.
In order to serve its purposes a guarantee mechanism should be set up in a
manner that reflects the state of technology and the expected trend of development.
A brief review of the technology will be given later in the paper together with an
assessment as to how stringent the warranties should be for any given stage of waste
management and disposal áctivity. The government should be well aware of its
responsibilities, which should be expressed in explicit terms rather than inferred
from immediate exigencies. This is also necessary for maintaining public confidence
in nuclear power. On the other hand, it is not in the best interest of society to
punish nuclear power, or any other form of power production for that matter,
with unreasonable regulations or penalties. Therefore the assurance schemes must
take into account the status of technology.
The viewpoints and the general reasoning given in the paper are relevant for
a country which is largely dependent on foreign technology. This implies that
nuclear power is not given any credit for being a diversifying factor or for increasing
the assurance of energy supply. These factors will naturally be considered on a
national level in any country and this could well result in somewhat more lenient
requirements than those to be presented here. It should be pointed out that this
paper is not a description of an existing system although in a number of instances
it is parallel to the proposals being considered or even adopted in Finland [ 1 ].
2. GENERAL REQUIREMENTS
The financial assurances and the institutional arrangements are to be derived
from certain general requirements of the following nature:
(1) The responsibility for safe disposition of spent fuel and all the wastes
generated must be uniquely defined and assigned to a specified holder
throughout the fuel cycle and at all stages of the waste management and
disposal operations.
(2) The organization or party having the title must exhibit the degree of
longevity which is compatible with the duration and extent of the
responsibility as measured in terms of the technical characteristics of the
spent fuel and wastes.
IAEA-SM-243/119 5
(3) Financial assurances must be provided to cover all expected future costs
and contingencies. The necessary escrow funds are to be accumulated
during the active lifetime of the reactor or the facility concerned. Any
additional costs are passed on to the utility industry.
The foregoing principles should be applied to all the wastes from the fuel
cycle, i.e.
(a) uranium (thorium) mines and mill tailings,
(b) front-end processing wastes, including wastes from conversion and
enrichment as well as from fuel fabrication,
(c) enrichment tails to the extent they are not exploited as blanket material,
(d) reactor wastes,
(e) spent fuel, including the stages of cooling pond storage, interim storage
and reprocessing or fuel element encapsulation and final disposal,
(f) reprocessing wastes, whether low, medium or high-level waste and
including liquid storage, solidification and conditioning, interim storage,
packaging and final disposal,
(g) decommissioning of reactors and other fuel cycle facilities, handling
and disposal of the decommissioning wastes.
The issues related to mill tailings or enrichment tails differ technically from
the other waste types. Items (a) and (c) appear just for the sake of completeness
and will not be discussed subsequently. One is just reminded of the fact that
further technical development is foreseen at least as regards mill tailings, which are
eventually likely to receive more attention even from the institutional point of view.
The remaining waste types in the list will be discussed following a conventional
categorization based on the level of radioactivity and half-life.
3. PRACTICAL ISSUES
The set-up can be formulated in a more practical manner than expressed in
the general requirements above. This reformulation also reveals immediately the
logical coupling between the technology and the requirements of a financial and
institutional kind.
The institutional requirement [2] can be stated in terms of considering the
role of the government or a governmental agency in carrying the responsibility.
To which waste categories and to which stages of the waste management and
disposal operations should the government liability be extended or limited? It is
an unavoidable fact that a government having granted a licence for the earlier
stages of a power programme has already tacitly assumed an implicit share of the
6 JAUHO and SILVENNOINEN
А В С D is c o u n t e d c o s t
F I G . I . A p r o b a b i l i t y d e n s i t y f u n c t io n o f th e d i s c o u n te d w a s te m a n a g e m e n t c o s t .
responsibility for the wastes and committed the country to allow waste generation
within its territory.
The requirement of financial assurances, evident as it is, becomes more intricate
because of the uncertainties embedded in the technology and, consequently, in
the cost. This is illustrated in Fig. 1. The extent of the financial liability should be
determined by the discounted present value of the cost of the storage, handling,
conditioning and disposal of the amount of waste generated or by the present
value of the decommissioning cost including the handling and disposal of the
decommissioning waste. Due to the uncertainties involved, the discounted cost (DC)
would be given in terms of a probability density function as is depicted in Fig.l.
The expected value В of the discounted unit cost DC would be a natural
measure of the liability. Once the DC-curve can be determined with an adequate
degree of certainty, the financial responsibility of the operator can be extended
to a value O B which would then already include an insurance type charge. In all
cases the government shall assume the risk of exceeding the cost C. In view of the
fact that unforeseen incidents can occur it should be emphasized that the
governmental liability may have to be activated in reality and therefore it should not
be taken as just another theoretical nuance.
This approach naturally implies the judgement that the costs can be evaluated
on a meaningful basis. Unfortunately, the uncertainties related to the future cost
trends remain too large to allow the exercise of Fig. 1 in any great detail. The
uncertainties in technology will subsequently be interpreted as a range in the cost.
An assumption of a constant probability will then transform the area under the
curve of Fig. 1 into a rectangle.
These two considerations pertain to the public interest. It goes without saying
that a scientist would grasp practical issues from a different angle. The other papers
of this conference will provide a far more detailed description of the status
of technology, whereas the review to be given below addresses itself mainly to the
two general issues.
IAEA-SM-243/119 7
Even for a given waste category the sophistication of the waste treatment
varies according to the expected mode of disposal. The conditioning technology
of waste is likely to be influenced by the experiences in the disposal stage, and in
deep underground disposal in particular. At a mature state of technology the
optimization aspects will inevitably attract more emphasis, while at the present
time the practices are very much concentrated around safety.
4.1. Treatment of low and medium-level wastes
As far as low-level solid wastes are concerned the problem is economic in
nature due to the large volumes of wastes generated. Volume reduction either by
compression or by combustion appears to be favoured instead of direct packaging
in drums. The volume of low-level liquid wastes can generally be reduced already
at the design stages of the plants by providing for recycling of the process water
whenever practical. Some of the low-level liquid wastes carry low contamination
and can be released.
The most effective method of volume reduction for both low and medium-
level liquid wastes is evaporation. The reduction factor varies typically from
5 to 10. The concentrates are immobilized either in concrete or bitumen.
Further development work is under way [2,3] and some new methods involve the
use of plastic resins.
A particular concern in the low-level regime is that of the long-lived alpha-
contaminated waste typically generated at fuel fabrication plants. For the
combustible fractions the ash residue is immobilized arid canned for disposal.
A similar example in the medium-level domain is the treatment of cladding
hulls and adjoining solid parts which may carry entrained fuel. These wastes can
be pressed mechanically to a more compact form or they can be melted. After
the reduction of volume the residue is canned.
The conditioning technology of low and medium-level wastes, whether short
or long-lived, may be considered well advanced in the sense that a government
involvement is not a necessity in any other capacity but the regular licensing
functions. A substantial development effort will still be made in the field but the
critical problems are essentially resolved.
4.2. Treatment of high-level waste
The evaporated concentrates of nitrate solutions of fission products and
actinides (neptunium, americium, curium) from reprocessing are stored typically
for five to ten years in cooled stainless steel containers. The treatment continues
with direct solidification or may include the separation of the actinides prior to
solidification.
4. TECHNOLOGY OF WASTE CONDITIONING
8 JAUHO and SILVENNOINEN
The transmutation of the separated actinides would yield fission products
with much shorter half-lives. To achieve a state of practical demonstration will
require much additional effort and these techniques are by no means mature at
the present time.
The immobilization of the aged high-level liquid waste is based on calcination
and vitrification in boron-silicate glasses. The French AVM process is operating
on an industrial scale. The technology is still to be proven in the case of reprocessing
wastes from high bumup LWR fuel. The fission product concentration and the
diameter of glass cylinders are free parameters in addition to the detailed
composition of glass. The glass itself must exhibit mechanical and chemical integrity
against radiation- and temperature-induced effects as well as against leaching. The
materials of encapsulation of the glass cylinders require careful consideration and
the requirements depend on the type of the repository used for disposal. The
cooling period before encapsulation or disposal will evidently become a major
optimization parameter. More recent suggestions for immobilizing radioactive waste
include a melted mixture of the waste and certain minerals which is left to
crystallize into a synthetic igneous rock [4]. One could only wish that new
methods would prove satisfactory and that their inventors would not unduly attack
the application of the other methods.
4.3. Encapsulation of spent fuel
Only few methods have been proposed for immobilization of spent fuel [5].
These conceptual schemes involve a period of storage after which the assemblies
or the pins would be packaged in metal canisters. Much development work would
be required to verify the suggested or other methods.
4.4. Decommissioning wastes
The decommissioning alternatives span a whole range from mothballing and
entombment to complete dismantling [6]. No firm course of suggested action has
yet evolved. The volume of the decommissioning wastes as well as the associated
costs vary substantially. The use of the same area for siting of subsequent reactors
or facilities would make dismantling unnecessary and some mode of entombment
would do.
5. DISPOSAL TECHNOLOGY
The options of waste disposal are commonly studied in the perspective of the
duration of hazard. The major distinction would be made between short-lived and
long-lived waste [2] even if in some cases all the waste would be disposed of in the
same way.
IAEA-SM-243/119 9
5.1. Shallow land burial
Shallow land burial represents perhaps the method that would be most widely
available for the disposal of short-lived wastes. The disposal sites will be subject
to surveillance and therefore the disposal strategy would include an evaluation of
the time period over which the site is to be operated.
As to the level of radioactivity to be permitted in the shallow land disposal,
uniform conventions are unlikely to evolve even if in some cases a limit of 10 nCi/g
for transuranium nuclides has been reported [2]. In constrast to sea dumping,
shallow land burial is a confined operation which at least provides an easy possibility
to control the migration of radionuclides. For some countries, of course, shallow
land disposal would become expensive or is not applicable due to the lack of an
appropriate site.
A plausible alternative to shallow land disposal would be emplacement in
existing natural caverns or old mines. Depending on local conditions, however,
these may not often assure appropriate isolation from groundwater flow. Large-
scale experience is available from the operation of the Asse salt mine with the
disposal of low and medium-level wastes. This experience is already exploited in
the design of actual repositories.
Although shallow land burial can be regarded as a well evaluated operation
which would not produce excessive risks, government involvement can hardly be
limited to licensing and site approval, especially at the present time when experience
with long-term surveillance is still limited.
5.2. Deep underground repositories
Final disposal of high-level and alpha-bearing waste or spent fuel is a field
where recent experiences have shown that public acceptance tends to require ‘
solutions to the problems at an earlier stage than waste management programmes
are geared to provide. It is difficult, indeed, to judge the conceptual designs,
whether they are devised for the purposes of supporting the licensing of reactors
and facilities at other stages of the fuel cycle or whether they already represent
engineering design practices. It is obvious that experimental programmes and
demonstration plants will eventually close some of the gap.
Different host formations will set the limit of the maximum permissible design
value for heat generation and the groundwater permeability will determine the
admissible leaching rate of packaging materials. Given these values the engineering
design can be conducted and pathway analyses can be carried out. For a clean-cut
engineering project the premises would be excellent. In reality, however, the
campaign has only started. From the viewpoint of accumulating funds for the
future disposal operations one should include warranties for prolonged delays.
Dry salt formations, embedded or domal, have been studied perhaps more
intensively than other host media. Substantial effort has been directed to studies
10 JAUHO and SILVENNOINEN
of hard rock, most notably in Sweden. Argillaceous sediments are also an important
option. In many countries one single candidate has a prominent role due to its
predominant occurrence in the area concerned.
Supporting R&D work has been structured in all related sciences. Besides the
investigations on the basic isolation of the waste from the geological strata the
research covers a broad area in geology from fault detection and evaluation to
geomorphic changes and earthquakes. The hydrology-oriented studies extend to
effects from extreme phenomena such as glaciers.
Institutional mechanisms to lessen the burden on the siting and construction
of repositories can be envisaged but no actual scheme is likely to be implemented
in the near future. The party holding responsibility for a repository should have
an inherent degree of longevity which seems to be inconceivable for any other
organization but one associated directly with a national government. The countries
with nuclear power programmes have to be prepared to take care of the spent fuel
discharged from the reactors. They also have to assume the responsibility for the.
reprocessing wastes which may be returned back to the country. Riskwise these
transports may entail a hazard greater than that entailed in retaining the reprocessing
wastes in a repository close to the reprocessing or vitrification plants. This is just
an example of the overriding political facts that upset not only the safety but the
economics as well.
5.3. Application of safeguards to repositories
The prospect of withdrawing spent fuel from the fuel cycle and disposing of
it brings a new dimension to waste disposal. As far as diversion of uranium or
plutonium to explosives is concerned the recycle strategies would place a relatively
low importance on safeguards at the disposal stage. To control the material
composition in the waste stream from reprocessing would be adequate. The spent
fuel repository in any nuclear programme based on once-through fuel utilization
will carry an attractive plutonium inventory.
Once one is to move from a conceptual stage to an engineering design the
safeguardability will emerge as one of the criteria effecting the transfer logistics and
surveillance instrumentation. IAEA will no doubt get involved in extending its
functions. The extreme long-term aspect invites one to speculate with different
options. For example, having the records kept would serve the purposes of safety
whereas having them disappear entirely would protect against malevolent diversion.
6. OPERATIONAL RESPONSIBILITY
The cursory review of technology given in the preceding chapters was to serve
the purpose of showing the extent to which the government has to assume the
IAEA-SM-243/119 11
responsibility for waste management and disposal. Since the financial obligations
rest with the waste producer it is proposed that the operational and financial
responsibilities are defined as separate liabilities. The operational responsibility
comprises the title and custody of the spent fuel and of the radioactive waste as
well as the responsibility and title of the reactors and facilities to be decommissioned.
The financial responsibility means the liability to cover the cost of carrying the
operational responsibility.
It is clear that in many countries where nuclear power production is the
business of a governmental agency the government is already adequately involved
as long as waste management is seen as an inseparable liability. On the other
hand, in a market economy there may be investor-owned utilities or cases where
government is a or the stockholder of the nuclear utility. The utility is governed
by the same company laws as any other business and clearly the necessary
government involvement is not inherently guaranteed.
A scheme of transferring the operational responsibility gradually from the
utility or plant operator to the government is presented in Table I. The base case
is defined for the subsequent discussion where a case study is performed to
illustrate a reserve fund system. Table I covers only the LWR cycles with recycle (R)
and once-through (O) options. This table is made consistent with the view given
earlier,.on the state of technology. In the case of certain items the operational
responsibility is retained in the private sector rather than transferred to the
government. For example, the construction and emplacement operation of the final
repository can possibly be carried out by a non-governmental operator whereas
the surveillance of an unactivated repository should always be left to the government.
7. A RESERVE FUND SYSTEM
It is obvious that an approach to provide financial assurances to cover future
costs and contingencies would be to establish special waste management funds [2].
This approach is more flexible than imposing an additional tax on electricity from
nuclear power plants. A fund system could be more proper to accommodate changes
in the cost evaluations and it would provide a more prompt response mechanism.
Recognizing that only a part of the waste management and disposal operations
will be in the custody of the government, it would be consistent to have a strictly
government-controlled fund limited to those operations where the government
is to assume the operational responsibility. Parallel to this the utility would
establish a special reserve fund to cover the liabilities over those stages where the
utility itself will carry the operational responsibility. This approach would not
absorb excessive contributions from the utility such that the utility would lose the
means to have any control and decision power over the reinvestments of the
accumulated capital.
12 JAUHO and SILVENNOINEN
TABLE I. OPERATIONAL RESPONSIBILITY IN WASTE MANAGEMENT
AND DISPOSAL
Stage of activity Fuel Base Altercycle case native
Low and medium-level waste
Treatment 0,R UTI
Disposal 0,R UTI GOV
Surveillance 0,R GOV
Spent fuel
Cooling pond storage 0,R UTI
Interim storage 0,R UTI GOV
Encapsulation 0 UTI GOV
Reprocessing R UTI
Treatment of high-level waste
Liquid storage R UTIVitrification R UTI
Storage of vitrified waste R UTI GOV
Deep underground repository
Construction 0,R GOV UTIEmplacement operation 0,R GOV UTISurveillance O.R GOV
Decommissioning 0,R UTI GOV
Transportation O.R UTI
GOV - Government, UTI = Utility or plant operator, R = Recycle, О = Once-through.
The application of this scheme to spent fuel is illustrated in Fig.2 in the base
case of Table I where the governmental custody is limited to the construction,
operation and surveillance of the final repository. A warranty fund is established
under the auspices of the government and with absolute governmental control.
The utility is obliged by the operating permit to place annually bonds with the
fund, the amount being equivalent to the present value of the estimated repository
costs as counted per the amount of the waste arising over the year concerned and
IAEA-SM-243/119 13
— — ^ Executive or institutional control Financial obligations
F Ï G .2 . I n s t i t u t i o n a l a n d f in a n c ia l a r r a n g e m e n ts in a r e f e r e n c e c a s e .
including the proportional part of the decommissioning cost and of the repository
costs of the plant decommissioning wastes. The payment will be waived only for
the part of the wastes that is permanently transferred to another country under
valid governmental agreements.
In addition to the escrow payments levied by the public fund the utility is
obliged to reserve funds within the company up to the amount corresponding to
the estimated future waste handling and disposal costs and excluding the final
repository. The extent of these funds will be annually approved by the authority
and these financial transfers must show up in the annual company balance.
Consequently, the utility is prevented from converting this part of its ostensible
profits to dividends. These funds are, however, a form of risk capital since the
utility itself is allowed to invest and use the money. The risk involved can be
removed by requiring the utility to present appropriate guarantees up to the amount
of its internal waste management funding reserves. These guarantees can be
presented as external assets or as capital investments apart from the nuclear power
plant and ancillary equipment.
TABLE II. UNIT COSTS OF LWR BACK-END FUEL CYCLE SERVICES3
14 JAUHO and SILVENNOINEN
Cost itemUnit cost
4% 10%
Reactor cooling pond $/kg-a 1-1.5 1-2
Interim storage $/kg-a 3-11 5-18
Shipments of spent fuel
reactor - interim storage $/kg 10-15 10-15
interim storage — disposal S/kg 28-32 28-32
interim storage — reprocessing S/kg 32-25 32-35
Shipment of HLW
reprocessing - storage $/kg SFb 10-15 10-15
storage — disposal $/kg SF 10-15 10-15
Encapsulation of spent fuel $/kg 15-23 18-28
Reprocessing $/kg SF 150-300 220-450
Vitrification and packaging $/kg SF 40-50 50-60
Storage of vitrified waste $/kg SFa 0.5 0.5
Disposal of spent fuel $/kg 60-66 74-80
Disposal of HLW $/kg SF 20-30 35-45
Surveillance $/kg-a 0.1 0.1
Uranium credit $/kg SF 140-215 140-215
Plutonium credit $/kg SF 97-240 97-240
a Estimates are valid for a large nuclear power system. b Spent fuel.
The arrangement described hinges upon tax laws and accounting practices
applied in any given country. It is assumed, of course, that no tax will be paid on
the utility revenues transferred to the funds.
While a charge per kW • h of electricity produced would make the system
somewhat easier to manage it would not then provide incentive to operate the plants
in a manner that would minimize the waste management expenses. Another
complication in the proposed scheme is the fact that inflation adjustments must
IAEA-SM-243/119 15
be implemented at times. If the waste management and disposal warranty were
specified in terms, of a certain percentage of the price of electricity the inflation
corrections would be made more automatic.
8. A PRACTICAL EXAMPLE
In order to illustrate how the system shown in Fig.2 would operate in practice
the annual contributions to the funds are calculated for a nominal LWR. The unit
cost data used are shown in Table II. The estimates are collected from recent
literature, compare Refs [7] and [8], for example. Each item is specified within a
range and the cost of financing is included at two interest rates of 4% and 10%.
No land cost is included in the figures.
To assess the warranty payment over one year (the year 0) the following time
schedule is assumed in the recycle case: cooling ponds storage 3 years, interim
storage 7 years, reprocessing during the 10th year and disposal of HLW during
year 20. In the once-through case the spent fuel would be disposed of during
year 20. The estimates of the present value of the total cost are given in Table III
for the two discount rates of 4% and 10%. Due to the ambiguity in the plutonium
credit the recycle option involves an assessment even with no plutonium credit.
The costs are split in two categories according to the base case defined in Table I.
Only the items spent fuel, reprocessing, treatment Of HLW and deep underground
repository are relevant for this example.
In almost all the cases the payments levied by the public fund would remain
lower than the internal reserve funding. The recycle cases exhibit much larger
variations. This effect is pinpointed if Pu credit is included. In the most favourable
cases there would not necessarily be any contribution to the utility fund. In fact,
the utility could speculate on a net present value credit up to 120 $/kg of spent fuel
at the 4% rate and 27 $/kg of spent fuel at 10%. The range 92—155 $/kg of spent
fuel is assumed in all the cases included in Table III.
In order to place the figures in perspective, the present value of the spent fuel cost
is expressed in terms of mills/kW-h in Table IV. A load factor of 70% was used.
The negative discounted value of spent fuel varies from 0.38 to 0.79 mills/kW-h
in the once-through case whereas in the recycle option the range is from a present
net value of 0.45 mills/kW-h to a cost of 0.85 mills/kW-h.
At this stage the government shall exert its own policy preferences. Any fixed
set of values in Table III (or in Table IV) helps the utility to select its course. In
case there is no policy preferred by the government the charges would be fixed
either at the most probable cost (A in Fig.l), at the expected cost (B) or at the
expected cost plus an insurance charge (C). Assuming a constant probability
distribution in Fig. 1 the discounted value of the expected cost would be
0.59 mills/kW-h for the once-through and 0.20 mills/kW h for the recycle option.
TABLE III. ESTIMATED PAYMENTS TO THE PUBLIC AND UTILITY FUNDS
16 JAUHO and SILVENNOINEN
R e fe re n c e case
P a y m e n t , $ /k g s p e n t fu e l
4% 10%
O n c e - th ro u g h
U ti l i ty fu n d 6 4 - 1 6 2 5 0 - 1 4 3
P u b lic f u n d 2 8 - 3 1 1 1 - 1 2
T o ta l 9 2 - 1 9 3 6 1 - 1 5 5
R e c y c le e x c lu d in g P u c re d i t
U t i l i ty f u n d 4 5 - 2 6 0 6 8 —2 4 9
P u b lic fu n d 1 0 - 1 2 5 - 7
T o ta l 5 5 - 2 7 2 7 3 - 2 5 6
R e c y c le w ith P u c re d i t
U t i l i ty f u n d 0 - 1 9 4 0 - 2 1 1
P u b lic fu n d 1 0 - 1 2 5 - 7
T o ta l 1 0 - 2 0 6 5 - 2 1 8
9. CONCLUDING REMARKS
In spite of the uncertainties involved, or in fact because of them, it would be
necessary from the point of view of public interest to establish warranty
arrangements to provide adequate financial and institutional means for the long
term management and disposal of radioactive wastes. The implementation of
these arrangements would place nuclear energy in a position where all the associated
costs would be included in the investment decisions. Consequently, the utilities .
would have a more objective basis in selecting the future modes of power production.
Waste management and disposal options for the waste generated can be
assessed at present with the degree of confidence that is required in a tentative
analysis of the proposed warranty arrangements. The accuracy would improve in
time and the establishment of the funds should be flexible enough to allow for
appropriate adjustments.
The back-end of the fuel cycle will influence the choice of different fuel cycle
options only partially. The considerations of resource availability will exert far
more influence on the decisions. The warranty requirements can, however,
be used by the government to guide fuel cycle policies. The uncertainties concerning
IAEA-SM-243/119
TABLE IV. DISCOUNTED COST OF SPENT FUEL MANAGEMENT
17
R e fe re n c e case C o s t, m ills /k W -h
4% d is c o u n t ra te
O n c e - th ro u g h 0 . 3 8 - 0 . 7 9
R e c y c le w ith P u c re d i t ( - 0 .4 5 ) —0 .8 5
the future cost, and the plutonium credit in particular, imply a wider range of cost
variation in the recycle option than in the once-through cases. In a large national
system with processing facilities of its own the policy-makers can exploit the
warranty funding fairly efficiently. In small countries depending on foreign services
the warranties should be determined in a conservative manner which would
emphasize the safety and security of services available.
ACKNOWLEDGEMENT
The authors are indebted to Mr. J. Vira for his assistance in carrying out the
computations.
REFERENCES
[1 ] A rra n g e m e n t o f N u c le a r W aste M a n a g e m e n t in F in la n d , M in is try o f T r a d e a n d In d u s t ry , S e rie s B :4 , H e ls in k i ( 1 9 7 8 ) .
[2 ] O b je c t iv e s , C o n c e p ts a n d S tr a te g ie s f o r t h e M a n a g e m e n t o f R a d io a c t iv e W as te A ris in g
f r o m N u c le a r P o w e r P ro g ra m m e s , O E C D /N E A , P a ris ( 1 9 7 7 ) .[3 ] D E J O N G H E , P ., “ A ris in g a n d m a n a g e m e n t o f n u c le a r w a s te s ” , T o p ic a l M e e tin g o n N u c le a r
P o w e r R e a c to r S a fe ty , B ru sse ls (1 9 7 8 ) .
[4 ] R IN G W O O D , A .E ., S a fe D is p o sa l o f H ig h L ev e l N u c le a r S p e c to r W aste s , A u s t ra lia n
N a tio n a l U n iv e rs ity P ress, C a n b e r ra ( 1 9 7 8 ) .[5 ] F in a l S to ra g e o f S p e n t F u e l , K B S , S to c k h o lm ( 1 9 7 8 ) ( in S w e d is h ) .
[6 ] B E R N E R O , R .M ., C O N T I, E .F . , “ D e v e lo p m e n t o f U n ite d S ta te s p o l ic y a n d s ta n d a rd s fo r
d e c o m m is s io n in g n u c le a r f a c i l i t ie s ” , D e c o m m is s io n in g o f N u c le a r F a c i l i t ie s (P ro c . S y m p .
V ie n n a , 1 9 7 8 ) , IA E A , V ie n n a ( 1 9 7 9 ) 11 .
[7 ] C h arg e f o r S p e n t F u e l S to ra g e , d r a f t e n v iro n m e n ta l im p a c t s t a t e m e n t , D O E /E IS -0 0 4 1 -D ,
U S D e p a r tm e n t o f E n e rg y , W a s h in g to n D C ( 1 9 7 8 ) .[8 ] R e g io n a l F u e l C y c le C e n tre s , V o l. I I , IA E A , V ie n n a (1 9 7 7 ) .
18 JAUHO and SILVENNOINEN
DISCUSSION
G. ROCHLIN: In your paper you indicate that money held by the utility in
escrow will be invested. Is this an economically efficient use of the money as
compared to self-investment by the utility in its own capitalization? What could
a utility invest in that is more secure than itself?
P. JAUHO: According to the suggested scheme the money in the escrow
fund is self-invested. The Ministry of Trade and Industry decides on the amount to
be kept as a reserve within the company and on the securities (e.g. letters of
guarantee) that may be required. Among the reasons for establishing a company
fund are the following: it would improve flexibility, provide an incentive to save
on costs in the conditioning of wastes, increase the financial stability of the
operator, make possible the allocation of costs to the waste producer at the time
the wastes are generated and facilitate inclusion of total waste-handling costs in the
price of electricity from the very beginning of plant operation.
G. STOTT: I refer to Fig.2 in your paper and the suggestion that only part
of the waste disposal management should be the responsibility of the government.
Is there not a danger that in these circumstances the utility will wish to determine
the manner in which the disposal is made, the rate of disposal and the general
management policy?
P. JAUHO: Under the licensing procedure, the safety authorities require a
firm plan for waste handling and management. This plan is approved by the
licensing authority (in Finland, the Ministry of Trade and Industry). In addition,
the Ministry makes rules for funding to cover future costs and approves securities
for the escrow fund.
IAEA-SM-243/77
THE USA’S PROGRAMME FOR
DISPOSAL OF RADIOACTIVE WASTES
J.M. BATCH
Battelle Memorial Institute,
Columbus, Ohio
C.A. HEATH
United States Department of Energy,
Washington, DC,
United States of America
Abstract
T H E U S A ’S P R O G R A M M E F O R D IS P O S A L O F R A D IO A C T IV E W A S T E S .
A fu ll-sc a le re v ie w o f th e U S A ’s w a s te m a n a g e m e n t p ro g ra m m e w a s r e c e n t ly c o n d u c te d
a t th e d i r e c t io n o f P re s id e n t C a r te r . R e c o m m e n d a t io n s m a d e as a r e s u l t o f th is re v ie w b y a n
In te ra g e n c y R ev ie w G ro u p o f th e U n i te d S ta te s G o v e rn m e n t h a v e e s ta b l is h e d th e b a sic e le m e n ts o f t h e U S p ro g r a m m e f o r d is p o s a l o f h ig h -le v e l ra d io a c t iv e w a s te s . T h e U S p r o
g ra m m e w ill b e b a se d o n th e a s s u m p t io n t h a t t h e f i r s t d is p o sa l fa c il i t ie s w ill b e m in e d
re p o s i to r ie s . L im ite d in v e s t ig a tio n s o f d is p o sa l in d e e p o c e a n s e d im e n ts a n d in v e ry d e e p h o le s w ill b e c o n t in u e d to a llo w t h e i r e v a lu a t io n as b a c k -u p c o m p e t i to r s . T h e p ro g ra m m e o f
in v e s t ig a t io n o f c a n d id a te s i te s w ill b e e x p a n d e d to c o n s id e r a g re a te r v a r ie ty o f g eo lo g ic
m e d ia . G re a te r e m p h a s is w ill a lso b e g iv e n t o d e v e lo p m e n t o f f u n d a m e n ta l s c ie n tif ic in f o r m a t io n o n th e e x p e c te d p e r f o r m a n c e o f r e p o s i to r y s y s te m s . N o t y e t re s o lv e d is th e
s c h e d u le o n w h ic h th e f i r s t o p e r a t in g r e p o s i to r y w o u ld b e l ic e n se d a n d c o n s t r u c te d .
A re v ie w o f c u r r e n t a n d p la n n e d a c t iv i t ie s in t h e U S p ro g r a m m e is p re s e n te d . E x p e r im e n ta l
te s t in g w ith e n c a p s u la te d s p e n t fu e ls is p la n n e d in fa c il i t ie s c u r r e n t ly u n d e r c o n s t r u c t io n .T he e x p lo ra tio n p ro g ram m e c u r re n tly u n d e r w ay in c lu d es fie ld te s tin g in sa lt d o m es in th e s o u th e a s te r n U S A , in b e d d e d s a lt f o r m a t io n s a n d in d e e p b a s a lt f lo w s . G e n e r ic te s t s a n d
p re l im in a ry s c re e n in g o f p o te n t i a l s i te s in o th e r m e d ia s u c h a s g ra n i te fo r m a t io n s a re b e in g
c o n d u c te d o r a re p la n n e d t o s ta r t s h o r t ly .
The programme to provide for the safe permanent disposal of radioactive
materials in the USA was greatly expanded in 1976 both in basic R&D and
exploration for potential sites. The results from the significant acceleration
of the scientific programme have recently been applied in an extensive evaluation
of the US policy of radioactive waste disposal in a programme initiated by
President Carter in April 1977. This paper describes the results of this extensive
review and summarizes the scientific status of the programme in the USA.
The policy review called for by President Carter resulted in the issuance of a
task force report by the Department of Energy in March 1978. Following the
issuance of this report, the President established an Interagency Review Group (IRG)
19
20 BATCH and HEATH
made up of all agencies within the US Government having some responsibility or
interest in the subject of radioactive waste disposal. The report of this Interagency
Review Group was submitted to the public for comment prior to its final
compilation for submission to the President of the United States. The final
recommendations on the national strategy for radioactive waste disposal were
submitted to the President early in 1979.
The National Environmental Policy Act (NEPA) requires that major decisions
of the US Government be supported by analysis of the environmental impacts of
the various alternatives prior to the decision. A “Generic Environmental Impact
Statement on the Management of Commercially Generated Radioactive Wastes”
was also issued for public comment in April 1979. This document provides the
necessary analysis of environmental factors to allow the formal adoption of several
recommendations made by the Interagency Review Group.
Key elements of IRG recommendations for the future conduct of the high-
level waste disposal programme may be summarized as follows:
• Near-term programme activities should be based on the assumption that the
first disposal facilities will be mined repositories. The nearer-term alternative
approaches of disposal in deep ocean sediments or very deep holes should be
funded to allow their adequate evaluation as competitors.
• Near-term R&D and site characterization programmes should be designed so that,
at the earliest date feasible, sites selected for location of a repository can be
chosen from among a set with a variety of potential host rock and geohydrologi-
cal characteristics. To accomplish this, R&D on several potential emplacement
media and site characterization work on a variety of geologic environments
should be increased promptly.
• A number of potential sites in a variety of geologic environments should be
identified and early action should be taken to reserve the option to use them
if needed at an appropriate time. In order to avoid working toward and
ultimatëly having a single national repository, near-term actions should create
the option to have at least two (and possibly three) repositories become
operational within this century, ideally, and insofar as technical and other
considerations permit, in different regions of the country.
• Construction and operation of a repository should proceed on a stepwise basis
and initial emplacement of waste in at least the first repository should be
planned to proceed on a technically conservative basis and permit retrievability
of the waste for some initial period of time.
The IRG technical findings concerning the geologic repository programme
were:
• A systems approach should be used to select the geologic environment,
repository site, and waste form.
IAEA-SM-243/77 21
• Overall scientific and technological knowledge is adequate to proceed with
region selection and site characterization, despite the limitations in our current
knowledge and modelling capability.
• Detailed studies of specific, potential repository sites in different geologic
environments should begin immediately.
• The actinide activity in transuranium-contaminated wastes and high-level
wastes suggests that both waste types present problems of comparable
magnitude for the very long term (i.e. greater than a thousand years).
• The degree of long-term isolation provided by a repository, viewed as a system,
and the effects of changes in repository design, geology, climate, and human
activities on the public health and safety can only be assessed through
analytical modelling.
• The effects of future human activity must be evaluated more carefully.
One issue concerning the establishment of high-level waste repositories which
was discussed in the IRG report remained unresolved at the time of preparation
of this paper. Still to be referred to the President of the United States for
determination of an Administration policy was the question as to how many
alternative sites should be evaluated before an attempt is made to seek licensing
approval to construct a repository. Some took the position that construction of
a repository should proceed as soon as one or more sites are identified as a result
of ongoing exploration in a few geologic media. Others argued that examination of
sites in a far wider number of geologic environments will be required to allow
a higher probability of success in the construction of the repository. Arguments for
and against each position were discussed at length in the IRG reports. The impact
on the US programme would be to affect the initial date of repository operation
with the evaluation of more geologic environments resulting in a delay of about
four years.
While the US programme will follow the general thrust of the recommendations
of the IRG following final decisions by the President, a final version of the generic
environmental impact statement will be published after receiving comments from
the public and the conduct of public hearings on its contents later this year.
The completion of the publication of this generic environmental impact statement
will be required under the National Environmental Policy Act prior to formal
adoption of the programme strategy.
The extensive review of the entire programme coupled with the great increase
in the number of participants and scientific specialists participating has resulted in
some changes in the direction of the US programme. The fundamental change
has been the recognition of a need for greater emphasis to be placed on developing
a broad scientific base behind the concept of geologic waste disposal. The first draft
of a programme plan developed jointly by the United States Geological Survey
22 BATCH and HEATH
(USGS) and the United States Department of Energy (USDOE), issued in January
1979, discusses many of the scientific requirements. This plan, entitled “Earth
Science Technical Plan for Mined Geologic Disposal of Radioactive Waste”,
discusses fundamental earth science problems which need to be addressed prior to
the licensing of geologic disposal facilities.
The plan represents a first step in the direction of addressing the main
technical questions that pertain to the safe isolation of radioactive wastes and to
showing how current and future tasks relate to these questions. The objectives
of the joint exercise include the following:
• Present a systematic review and identification of earth-science issues and
recommend work that will support site selection, characterization, and
evaluation; repository design and operation; and short- and long-term risk
assessment.
• Inventory and classify current technical R&D to facilitate integration and
co-ordination of USDOE- and USGS-sponsored work.
• Identify R&D activities which should be more closely co-ordinated for timely
resolution of technical questions.
• Specify technical plans to resolve significant uncertainties regarding mined
repositories, by building on the IRG report and other earth-science reports.
• Provide information which can be used to develop broad estimates of the time
required for resolution of earth-science technical questions and to organize
the work and resources to determine a realistic schedule for developing a
repository.
• Recommend the preparation of review reports on key technical problems
and of summary documents that will aid in planning future R&D activities.
The January report represented the first necessary stage of evaluation prior to
providing detailed recommendations for modifying current programmes to improve
co-ordination and for initiating future research. A more complete report will be
issued this summer.
Turning now to specific technical activities in the US programme, significant
progress in the modelling of geosphere transport mechanisms has taken place in the
last several years. Geosphere transport models have been developed and applied to
candidate geologic systems in order to evaluate the potential for radionuclide
migration in these candidate systems. The application of this analysis has been and
will continue to be used in the site selection process. A paper submitted to this
Symposium by Brandstetter et al. (IAEA-SM-243/35) describes in more detail the
progress to date in developing these analytical tools.
Careful analysis has pointed to a need to place more emphasis on multiple
barriers of containment in the geologic disposal system. The previous philosophy
IAEA-SM-243/77 23
of geologic disposal in the USA tended to place great emphasis on the primary
barrier of the geologic host rock in preventing the escape of radionuclides to the
biosphere. With an appreciation of the uncertainty in predicting future geologic
behaviour and the complexity of geochemical mechanisms occurring beneath the
surface of the earth, a recognition came of the need to depend upon multiple
barriers to build a higher value of certainty into the isolation concept. The
conceptual design studies performed by KBS in Sweden aroused great interest in
the USA and have led to a greater emphasis on the engineered barriers in addition
to geologic barriers. The value of engineered barriers, however, has to be assessed
in the light of the additional reliability that can be added to the system by their
implementation.
While the various nations of the world study in more detail the possible
future systems and fuel cycles for the use of nuclear power, the possibility remains
that spent fuel from éxisting light-water reactors may be disposed of without further
reprocessing. Our commercial waste management programme has been structured
to allow for the possibility that such future disposal might be required. Therefore,
without any prediction that spent fuel in the future may require disposal, the
programme has been designed to allow for this possibility.
An active programme is under way to characterize the spent fuel from light-
water reactors and to gain experience in the handling and packaging of spent fuel
for possible geologic disposal. Summary highlights of this programme include
the following.
A large portion of the programme involves the experimental packaging of
spent fuel. A large hot-cell facility located on the Nevada Test Site is being used
to prepare experimental packages containing spent fuel for use in programme
designed to demonstrate dry surface storage and geologic disposal. A view of the
inside of this facility is shown in Fig. 1.
Early in 1979, three PWR fuel packages were completed and placed in dry
surface storage as a part of a demonstration programme. Later in 1979, a BWR
spent-fuel element will be packaged and also placed in dry surface storage.
Later in 1979, 14 spent-fuel packages will be prepared and placed under
ground in granite in the Climax Mine on the Nevada Test Site. Construction
activities in this facility, which is located at a depth of 1400 feet, are shown in Fig.2.
This demonstration will be followed by the fabrication of 22 more spent-fuel
packages for placement in the Near Surface Test Facility in basalt at Hanford.
While extensive scientific studies are being pursued concerning the fundamental
requirements of radioactive waste disposal in geologic media, efforts are also under
way in the area of exploration for potential sites, engineering and design studies for
construction of future repositories and their associated equipment. Within the last
year three detailed conceptual designs have been completed on key facilities that
will be required in the geologic disposal programme. These have been a fuel-
handling and packaging facility; a 2000-acre repository located in bedded salt
24 BATCH and HEATH
FIG .l. A view of the inside of the EMAD facility in Nevada where experimental work is
being performed to package spent fuel.
designed primarily for the receipt of packaged spent fuel; and a repository for the
receipt of high-level waste located in a salt dome. Typical dimensions of salt domes
as they exist in the southeastern USA were used for this latter study and projections
made of potential repository capacities in this type of formation. The combination
of waste forms and geologic media was not intended to designate any particular
choice for the programme. Rather, the combination of alternatives was selected in
order to provide maximum information for the conduct of the programme.
Geologic repository designs in other host rock types, such as basalt, are to be initiated
in the very near future.
A primary emphasis of the past several years of the programme in the USA
has been to evaluate the potential of salt deposits for placement of geologic
repositories. Significant work has been done in this area, highlighted by the
characterization of a site in southeast New Mexico which has been proposed for
the placement of a facility called the Waste Isolation Pilot Plant (WIPP). An aerial
view of the proposed site is shown in Fig.3.
IAEA-SM-243/77 25
FIG.2. A view of construction activities inside the granite facility located in Nevada.In early 1980, packaged spent-fuel elements will be placed here for an experimental test.
The primary function of WIPP will be the receipt of transuranium-contaminated
(TRU) wastes, but it has also been proposed to incorporate what is known as an
intermediate-scale facility for the experimental placement of up to 1000 packaged
fuel elements from commercially operated LWR reactors. A paper submitted to
this Symposium by L. Hill (IAEA-SM-243/38) describes in detail the character
of this proposed site in southeast New Mexico.
Just as the recent review of the waste management programme highlighted the
requirement for a broader scientific base to the concept of waste disposal in
geologic media, it was also apparent that the investigation programme should
address a wider variety of geologic environments. The Interagency Review Group
26 BATCH and HEATH
FIG.3. A n aerial view o f the proposed location o f the WIPP facility located in southeast N ew Mexico.
recommendations call for early action to identify potential sites in a variety of
geologic environments and, if necessary, to reserve them until a selection is made of
a site to develop a repository. High priority is, therefore, being placed on qualifying
sites in a number of areas. Figure 4 identifies the dates by which it is anticipated
that sites beyond the proposed site in New Mexico might be found suitable for
potential repository locations.
Greater emphasis is now being applied to investigations of several geologic
media and host rocks for potential development of geologic repositories. The site
characterization process is focusing not only on specific host media but also on
surrounding geologic systems. It is recognized that the groundwater flow and other
hydrologie features, the surrounding mineral types and the entire geologic basin
system must be investigated in order to characterize any site as a potential geologic
disposal facility.
The unshaded triangles designate a point at which a number of specific sites
will be identified in each formation for more detailed examination. The shaded
triángle represents a point by which a site might be qualified to suffice for
us to be able to support an application to the licensing authorities.
IAEA-SM-243/77 27
FISCAL YEARS 1979 1980 1981 1982 1983 1984
TERMINAL STORAGE ©II II II
V1. GEOLOGIC SITE SELECTION
• SALTV V
G U LF DOMES
BEDDED S A LT - 1
BEDDED S A L T - 2
V T V ▼
V ▼
• N O N -SALT
HAN FORD BASALT
NTS MEDIA
O THER REGIONS
V ▼V ▼
V ▼
V MULTIPLE SITE CANDIDATES IDENTIFIED
▼ GEOLOGIC SITE IDENTIFIED AS SUITABLE FOR REPOSITORY II
V ALTERNATIVE STRATEGIES FOR IN IT IA L SITE SELECTION
@ EARLY SALT AND DOE SITE BASALT OPTIONS
® EXPANDED SALT AND DOE SITE NON-SALT OPTIONS
© MULTIPLE GEOLOGIC MEDIA OPTIONS
FIG.4. The earliest possible tim e by which suitable sites fo r a geologic repository may be fo u n d is show n by the shaded triangles.
The work in salt formations includes characterization of bedded salt formations,
salt anticlines in bedded salt formations, and the salt domes in the southeastern USA.
Major salt deposits exist in several broad areas in the USA, as shown in Fig.5.
Preliminary investigations indicate that several states have geologic formations of
particular interest. These include the Gulf Coast Interior Salt Domes in Mississippi,
Louisiana and Texas, the Permian bedded salt basin in Texas, the Paradox bedded
salt .hasin in Utah, and the Salina bedded salt basin in Michigan, Ohio, and New York.
At the present time explorations are active in all of these areas except the Salina
bedded salt basin.
The steps for exploration are somewhat similar for each of these salt deposits.
First, a characterization study is performed for a total region. Based on this study,
specific areas are identified for further exploration. After further study of each area,
specific salt deposits are studied. Included are such things as aerial infra-red
surveys, groundwater mapping, geomechanics studies, gravity surveys, seismic
data, and core drilling.
In parallel with the geologic studies, extensive investigations of the surface
environment are performed. These and other studies are combined to make a
decision as to whether any particular site would qualify for a repository site.
28 BATCH and HEATH
REGIONS BEING IN V E S TIG A TE D FOR T E R M IN A L STORAGE OF R A D IO A C T IV E WASTES
FIG .5. Regions being investigated fo r technical storage o f radioactive wastes in the USA.
All of the salt deposits mentioned previously and other rock types and geologic
environments that will be evaluated will be examined in a similar fashion.
A large basaltic flow exists in the northwestern region of the USA, covering
several states. The Hanford reservation, which has been the location of work in the
area of nuclear energy for several decades, is centrally located in the midst of this
large basaltic flow. The previous use of this reservation makes it a logical place to
be examined for its potential as a repository site. The characterization of the Pasco
basin and its associated basaltic flows is the subject of the paper by R.A. Deju
(IAEA-SM-243/36).
Arid regions of the western USA cover very large areas of land which may be
suitable for repository siting. The great basin of Nevada and Utah is a hydrologically
closed system with no drainage to the oceans. The Nevada Test Site located in this
basin has also been the centre of extensive work associated with nuclear energy in
the past decades. This reservation contains several media of potential interest for
geologic disposal. Volcanic tuffs and shales are presently under investigation as
possible locations for future repository placement.
IAEA-SM-243/77 29
Work in granite in the USA has until now been associated mostly with generic
suitability studies. The co-operatively funded programme being conducted with
Sweden in the Stripa mine has been extremely valuable in providing more
information about the potential suitability of granite as a host rock for a geologic
repository. In addition, conveniently located granite workings at the Nevada Test
Site have allowed the conduct of parallel tests within the USA. Unfortunately,
the particular granite system which has been used is not suitable for location of a
geologic repository, primarily owing to other geologic factors in the area. As
previously described, a wider programme of geologic exploration will be carried
out and will include potential granite sites in the USA.
The USA’s programme as presented covers a very wide range of investigations
of both scientific interest and site-specific geologic media. The widespread review
of the waste management programme conducted at the direction of the President
of the United States revealed a growing consensus among the scientific community
that the safe and permanent disposal of radioactive wastes in geologic media can be
successfully achieved. However, it did indicate a need for a wider scientific
investigation in order to assure that significant questions that still remain can be
satisfactorily answered prior to the licensing and irretrievable placement of
radioactive materials in any geologic media. As a result of this Government-wide
review, however, the Department of Energy is receiving additional and necessary
support in the conduct of the programme for commercial waste management from
other agencies of the United States Government, whose co-operation will be required
for the successful implementation of the geologic disposal programme.
DISCUSSION
K. KÜHN: With regard to the status of the WIPP project, may I ask if a
decision has been taken by the White House? Is the project to be licensed or is it
to be handled as a military project which does not have to be licensed? If no
licensing procedure is applied, what will be the impact of this project?
C.A. HEATH: Up to 29 June 1979, when I left the USA, no final decision
concerning the status of the WIPP project had been reached. The Department of
Energy has proposed that this project should be expanded to include the placement
of some spent-fuel elements from commercial reactors and that the project should
be licensed by the regulatory authorities. However, Government funds cannot be
spent in the USA until specifically authorized by the United States Congress.
The Congressional Committee which originally approved the WIPP project does
not agree that the project should be made subject to regulation by the licensing
authorities since its primary purpose is disposal of transuranium-contaminated
wastes from military programmes. Until agreement is reached between the
Administration and the Congress on the WIPP project, its future is uncertain.
30 BATCH and HEATH
H. KRAUSE: Some years ago ERDA started a large programme on the disposal
of radioactive wastes. Is the programme you have just described a continuation of
that ERDA programme or is it a different one?
You mentioned that by the end of this century two or three disposal sites will
be in operation in the USA. Does this mean that you have a maximum of two or
three different geologic media? Will the fundamental investigations you referred to
concentrate on these candidate geologic formations or will they also cover other
potential formations?
C.A. HEATH: The programme now under way is a continuation of the one
announced by ERDA, but significant revisions have been made in the details of the
earlier project as a result of more recent studies and the review by the Interagency
Review Group.
The two to three facilities in different regions of the country represent an
objective or goal for the programme. These facilities do not have to be in different
geologic media although they may very well be.
The investigations under the USA’s programme will be site-specific studies in
particular geologies and also fundamental or generic studies concerning reactions
between various waste forms and geologic media.
H.O. BOHM: You mention in your paper that one of the findings of the
Interagency Review Group is that a systems approach should be used to select not
only the geologic environment and the repository site but also the waste form.
Does this mean that it is still an open question in the USA whether you will
reprocess the spent fuel or merely dispose of the unreprocessed LWR fuel elements?
If so, what, in your opinion, are the main criteria in the USA for reaching decisions
in these matters?
C.A. HEATH: The waste management programme in the USA is being
conducted to allow for the possibility either that spent fuel will have to be disposed
of or that reprocessing will be reinstituted and reprocessing wastes will need to be
disposed of. A decision as to whether reprocessing will take place is expected to be
made on issues other than those related to waste management. At present, we believe
that either waste type will be suitable for disposal.
H.O. BOHM: You presented some information about the United States
programme on the packaging of spent fuel. Does this programme deal only with
simple packaging of whole fuel elements or are you also considering the application
of some other types of conditioning to the spent-fuel elements which are to be
disposed of?
C.A. HEATH: The experimental programme referred to deals primarily with
a simple encapsulation of fuel elements in a steel container with only a helium
backfill. Additional work is planned to determine whether some form of
conditioning of the spent fuel will be required. Possible requirements for disassembly
of the fuel elements and release of the pressurized fission gases will be considered
as well as the possible need for other backfill materials in the canister.
IAEA-SM-243/1 IS
ПРО ГР А М М А И С С Л Е Д О В А ТЕ Л Ь С КИ Х РАБО Т
ПО О К О Н Ч А ТЕ Л Ь Н О М У ЗАХО РО НЕНИЮ
В Ы С О КО А КТИ В Н Ы Х О ТХО Д О В
В Г Л У Б О К И Е С ЛА Б О П РО Н И Ц А ЕМ Ы Е ГЕ О Л О ГИ Ч Е С К И Е
Ф О РМ АЦИИ
О. Л. КЕДРОВСКИЙ, Е.А. ЛЕОНОВ, Н.М. РОМАДИН, И.Ю.ШИШИЦ
Представлен М. К. Пименовым Государственный комитет
по использованию атомной энергии СССР,
Москва,
Союз Советских Социалистических Республик
Abstract- Аннотация
RESEARCH PROGRAMME ON THE DISPOSAL OF HIGH-LEVEL WASTES IN DEEP
GEOLOGICAL FORMATIONS WITH LOW PERMEABILITY.
From the point of view of reliability, the lithosphere is the most promising medium
for the disposal of radioactive wastes, and those parts of it which are of low permeability and
which would ensure that wastes were contained within a given area are of particular interest.
It is those parts of the lithosphere which have been taken as the basis for drawing up require
ments for disposal in artificial underground repositories, boreholes, disused or specially dug
mine shafts of different widths and designs. It is shown that in practice any type of disposal
requires the solution of a number of scientific and technical problems, the most important
of which concern the mining geology and mining technology profiles. The main tasks to bé
performed are as follows: determination of the parameters of the temperature field produced
by the build-up of heat from wastes and of the deformational stresses on rock of different
types with different structural and hydrogeological characteristics; evaluation of the
geochemical and physical changes which could be induced in different types of rock by high
pressure, high temperature and high radiation doses in the case of contact with high-level
wastes; estimation of the effect of placing wastes in the repository on the geological, hydro-
geological and filtration characteristics of the disposal sector and neighbouring areas;
elaboration of layouts, models, process equipment and design of repositories for different
mining geology and mining engineering conditions and for wastes of different chemical
composition, energy release, shape and consistency; establishment of mining geology, mining
engineering, health and other criteria for evaluating the suitability of disposal sectors and
surrounding areas; elaboration of requirements for prospecting and making inventories of
suitable sectors and areas for waste disposal; formulation of specifications for construction
materials; evaluation of the suitability of existing materials and development of new ones;
development of a surveillance system and of a method of sealing the repository, and so on.
Since these questions have not yet been studied in sufficient depth, they have been included
in a many-sided programme of scientific research and experimental design work consisting
of the following: theoretical studies, laboratory simulations, large-scale laboratory and field
experiments, planning and experimental disposal on an industrial scale of real wastes and
development of standard manuals and technical norms. The programme also includes studies
on the suitability of many types of rock, such as rock salt, clay, granite, porphyrite, diabase,
tufa and so on to serve as repositories.
31
32 КЕДРОВСКИЙ и др.
ПРОГРАММА ИССЛЕДОВАТЕЛЬСКИХ РАБОТ ПО ОКОНЧАТЕЛЬНОМУ ЗАХОРОНЕНИЮ ВЫСОКОАКТИВНЫХ ОТХОДОВ В ГЛУБОКИЕ СЛАБОПРОНИЦАЕМЫЕ ГЕОЛОГИЧЕСКИЕ ФОРМАЦИИ
Наиболее перспективной средой с точки зрения надежности захоронения радиоактивных отходов является литосфера. Особый интерес представляют слабопроницаемые ее участки, позволяющие локализовать отходы в пределах заданного контура. Применительно к таким участкам сформулированы требования, предъявляемые к захоронению в подземные искусственно создаваемые емкос- ти-хранилишд, в буровые скважины, в отработанные или специально создаваемые шахты и шахтные стволы-колодцы различной конструкции. Показано, что практическое осуществление любого варианта захоронения требует решения ряда научно-технических задач, главными из которых являются задачи горно-геологического и горнотехнического профиля. К числу основных задач отнесены: определение параметров наведенного разогревающимися отходами температурного поля и напряжен- но-деформированного состояния горного массива, представленного различными породами, при различной его структуре и гидрогеологии; оценка возможных геохимических и физических превращений различных горных пород в условиях повышенного давления, высокой температуры и большой радиационной нагрузки при контакте с высокоактивными отходами; установление влияния размещенных в хранилище отходов на геологию, гидрогеологию и фильтрационные свойства участка захоронения и соседних районов; разработка принципальных схем, моделей, технологических приемов, конструктивного оформления хранилищ в различных горно-геологических и горнотехнических условиях для различных по химическому составу, энерговыделению, форме и консистенции отходов ; установление горно-геологических, горнотехнических, санитарных и других критериев для оценки пригодности участков и районов захоронения; разработка требований на проведение изыскательских работ и инвентаризации пригодных участков и районов для захоронения отходов; формулирование требований к конструкционным материалам; оценка пригодности существующих и разработка новых материалов; разработка системы контроля и порядка консервации хранилищ и т.п. Недостаточная изученность затронутых вопросов обусловила включение их в разработанную комплексную программу научно-исследовательских и опытно-конструкторских работ, предусматривающую теоретические исследования, лабораторное моделирование, крупномасштабные стендовые и полевые испытания, проектирование и опытно-промышленное захоронение реальных отходов, разработку нормативных справочников и технологических регламентов. Программой предусмотрено изучение пригодности для захоронения многих горных пород, включая каменную соль, глину, гранит, порфирит, диабаз, туф и др.
Рост мощной ядерной энергетики и радиохимических производств неизбежно при
водит к наработке радиоактивных отходов. Наибольшую потенциальную опасность в
настоящее время и в будущем представляют высокоактивные и альфа-активные отходы,
требующие надежного их удаления из сферы жизнедеятельности с изоляцией до прекра
щения их вредного воздействия.
Наиболее перспективной по надежности захоронения высокоактивных долгожи
вущих отходов средой является литосфера. Особый интерес представляют слабопро
ницаемые ее участки, позволяющие осуществить вечную локализацию отходов в пре
делах заданного контура.
Захоронение высокоактивных отходов в слабопроницаемые геологические фор
мации должно удовлетворять следующим основным требованиям:
IAEA-SM-243/115 33
— обеспечение надежной изоляции размещенных в хранилище отходов от биосфе
ры в течение неограниченного периода времени;
— захоронение допустимо лишь в горные массивы, обладающие коэффициентом
фильтрации в пределах 1СГ5- 1СГ3м/сут или проницаемостью 10“5-10“3дарси;
— недопущение распространения радиоактивности в горном массиве (с газами
или водой) на расстоянии не более 400 м от контура выработок-хранилищ. Указанное
расстояние, как показали предварительные расчеты, основанные на решении фильтраци
онных задач для слабопроницаемых подверженных прогреву горных массивов, с уче
том сорбционных свойств горных пород, позволяет реализовать значительное число
вариантов захоронения отходов и не вызывает больших осложнений в части отчуждения
земельных участков, проведения тщательной геологической разведки и осуществления
контроля в процессе эксплуатации хранилища;
— зона горного массива, в которой допускается со временем появление следов
радиоактивности, вызванное диффузией или миграцией размещенных в подземных
емкостях отходов, должна быть полностью изъята из сферы жизнедеятельности чело
века с момента захоронения отходов на неограниченный период времени. Это — зона
локализации отходов с геометрическими параметрами, рассчитанными на основе фак
тических свойств горного массива, с учетом всех эффектов, сопровождающих вечное
захоронение реальных радиоактивных отходов ;
— емкости-хранилища должны располагаться ниже зоны свободного водообмена
с оставлением между этой зоной и контуром зоны локализации охранной породной тол
щи мощностью не менее 70-100 м. Эти цифры вытекают из решения задачи о проница
емости породного слоя при заданных перепаде давления, свойствах породы, времени
фильтрации и вязкости фильтрующей среды;
— уменьшение глубины размещения зоны локализации отходов допустимо лишь
на участках, представленных мощной толщей сухих водо-газоупорных пород, выходя
щей на поверхность земли;
— размещение хранилищ должно производиться на расстоянии не менее 1000 м
от зон тектонических нарушений;
— хранилища могут размещаться только в районах, в которых исключаются ин
тенсивные землетрясения и неотектонические преобразования горного массива;
— расположение хранилищ допустимо лишь за пределами месторождений полез
ных ископаемых, имеющих значение для промышленности в настоящее время и в перс
пективе;
— обеспечение агрегатного состояния захороненных отходов, исключающего их
распространение по трещинам и порам, за пределы зоны локализации;
— обеспечение долговременного контроля за температурой, давлением и уровнем
активности в хранилище и в прилегающем к нему горном массиве;
— места строительства хранилищ должны быть существенно удалены от подзем
ных сооружений и наземных объектов, не имеющих прямого отношения к радиохи
мическому производству и атомным электростанциям.
34 КЕДРОВСКИЙ н др.
В связи с изложенными требованиями возникла необходимость разработки тех
нологии захоронения отходов в слабопроницаемые геологические формации, основ
ными элементами которой являются:
— классификация способов захоронения предварительно отвержденных и отверж
даемых в горном массиве отходов ;
— параметрический ряд подземных хранилищ с их технической характеристикой
и принципиальной схемой расчета конструктивных элементов;
— совокупность критериев, методов и средств по изучению пригодности геоло
гических участков для строительства хранилищ, пригодных для вечного захоронения
отходов;
— система контроля обстановки в хранилище в процессе захоронения отходов и
консервации сооружения с регистрацией температуры, давления и уровня активности;
— система автоматизированного контроля за состоянием элементов хранилища
и горного массива, снабженная телеметрическими каналами передачи информации;
— совокупность методов и средств контроля за локализацией радиоактивности
в пределах отчужденного под захоронение геологического участка;
— система методов и средств регулирования термодинамических и физико-хи
мических процессов при захоронении отходов и консервации хранилища;
— совокупность методов и средств ликвидации аварийной обстановки в случае
ее возникновения;
— каталог механических и гидравлических средств доставки отходов в рабочее
пространство хранилища, в том числе дистанционных манипуляторов, позволяющих
осуществлять операции по перегрузке, размещению и фиксации отходов в выработках;
— комплекс дистанционного слежения за перегрузочными и доставочными операци
ями.
Технология захоронения отходов в слабопроницаемые формации не является изо -
лированной и тесно связана, с одной стороны, с технологией подготовки отходов к захо
ронению и их транспортировкой и, с другой стороны, с технологией переработки и обез
вреживания образующихся и выходящих из хранилищ парогазовых и газовых смесей.
Обе эти технологии играют весьма важное значение в общей проблеме обезвреживания
и удаления радиоактивных отходов из сферы деятельности человека, составляют само
стоятельную задачу и рассматриваются отдельно.
Работая над решением проблемы захоронения высокоактивных отходов в слабо
проницаемые горные массивы, нашими специалистами рассматриваются следующие
принципиальные схемы;
— захоронение в подземных искусственно создаваемых емкостях-хранилищах;
— захоронение в буровых скважинах;
— захоронение в отработанных или специально сооружаемых шахтах;
— захоронение в шахтных стволах-колодцах различной конструкции.
Разработка технологии захоронения по той или иной схеме требует решения
большого ряда научно-технических задач, главными из которых являются задачи горно
геологического и горнотехнического профиля. Это связано с тем, что высокоактивные
IAEA-SM-243/115 35
отходы, выделяющие большое количество тепла в течение нескольких сотен лет, обла
дающие химической и радиационной агрессивностью, могут существенно изменить ес
тественные свойства горного массива, подвергая его геохимическим превращениям и
разрушению. Указанные отходы отрицательно влияют не только на свойства горного
массива, но и на конструкционные материалы (тампонажный раствор, металл, бетон
и др.), что может привести к разгерметизации подземных хранилищ.
Большое влияние на эффективность захоронения оказывает результат выбора
района расположения хранилищ и геологической формации. Располагая хранилища
с отходами вблизи зон тектонических нарушений, в неоднородных трещиноватых мас
сивах, в породах с большим газовыделением при их прогреве и т.д., возникает потен
циальная опасность неконтролируемого распространения радиоактивности по массиву
и осложняется долголетнее обезвреживание парогазовой смеси, истекающей из очага
разогрева.
К числу основных горно-геологических и горнотехнических задач относятся:
— определение параметров наведенного разогревающимися отходами темпера
турного поля в горном массиве, представленного различными породами при различ
ной его структуре и гидрогеологии ;
— анализ картин напряженного состояния различных горных массивов вокруг
объекта захоронения при одновременном длительном действии внутреннего давления
и высоких температур;
— оценка возможных геохимических превращений различных горных пород,
оказавшихся в условиях избыточного давления, высокой температуры при контакте
с высокоактивными отходами;
— оценка возможных перемещений отходов в массиве за счет плавления пород
и миграции;
— изучение влияния высокотемпературного очага с избыточным давлением на
гидрогеологию участка захоронения и соседние районы;
— разработка принципиальных схем (моделей) захоронения отходов применитель
но к различным горно-геологическим и горнотехническим условиям и различным видам
отходов;
— определение минимально допустимых расстояний от хранилища до проницае
мых зон (фильтрующие горизонты, разломы, тектонические сдвиги и др.);
— разработка горно-геологических, горнотехнических и санитарных критериев
для оценки пригодности участков (районов) захоронения высокоактивных отходов
различного вида;
— разработка требований на проведение изыскательских работ по выбору участка
захоронения отходов;
— инвентаризация имеющихся потенциально пригодных районов для захоронения
отходов как вблизи источников образования отходов, так и в местах, которые можно
использовать для создания централизованных хранилищ;
— разработка системы многолетнего контроля за параметрами протекающих в
хранилище процессов (температура, давление, уровень активности, химический состав) ;
36 КЕДРОВСКИЙ и др.
— разработка системы многолетнего контроля за степенью локализации отходов
в заданном контуре;
— формулирование требований к конструкционным материалам и оценка степе
ни пригодности существующих материалов;
— разработка положения о порядке консервации хранилища;
— оценка влияния сейсмических импульсов и процессов геологической метаморфи-
зации на надежность локализации отходов.
Далеко не полный представленный перечень основных горно-геологических и гор
нотехнических задач свидетельствует об исключительной сложности проблемы захоро
нения в геологические формации, требующей для своего решения применения совре
менных методов исследований и специально разработанной аппаратуры.
В процессе исследований предстоит рассмотреть поведение в сложных условиях
многих горных пород, включая каменную соль, глины, граниты, порфириты, диаба
зы, туфы и др.
Сами породы и представленные ими горные массивы применительно к проблеме
захоронения в них разогревающихся радиоактивных и химически активных отходов
до последнего времени изучались недостаточно. Отсутствует общепризнанная методика
решения задачи о прочности реального горного массива при одновременном действии
на него горного давления, температурных нагрузок, внутреннего давления, химической
агрессии и высокой радиации.
Отмеченное остается в силе и по отношению к конструкционным материалам.
Методика и состав проводимых в настоящее время изыскательских работ не поз
воляют получить исходные данные в объеме, необходимом для научно обоснованного,
безопасного захоронения отходов.
Изложенное выше обусловило необходимость разработки комплексной програм
мы научно-исследовательских и опытно-конструкторских работ.
В соответствии с этой программой предусматриваются: выполнение теоретичес
ких исследований; моделирование в лабораторных условиях; создание крупномасш
табного стенда для изучения поведения горного массива при совместном действии всех
основных нагрузок; проведение опытов на прототипах, созданных у поверхности зем
ли и в действующих шахтах; разработка проектов опытно-промышленного захороне
ния реальных отходов и их реализация; конструирование и опробование автоматизиро
ванных систем контроля за поведением отходов в хранилище и состоянием горного мас
сива; разработка критериев по оценке пригодности различных участков для захороне
ния и требований на изыскательские работы.
Ниже излагаются основные положения программы и методика решения некото
рых задач.
Основной особенностью программы является то, что в ней заведомо не отдается
предпочтения определенной технологической схеме захоронения или ее модификации,
поскольку в настоящее время еще нельзя с полной определенностью назвать оптималь
ный уровень активности захораниваемых отходов и не представляется возможным су
дить о преимуществах вариантов захоронения без искусственного охлаждения по срав
IAEA-SM-243/115 37
нению с вариантами, требующими воздушного или водяного охлаждения. Это связано
с тем, что варианты, рассчитанные на полный отвод тепла, благодаря теплопроводности
горного массива, сопряжены с необходимостью большого рассредоточения отходов в
массиве и, следовательно, с большим объемом горных работ на единицу объема захо
раниваемых отходов. Кроме того, эти варианты исключают утилизацию тепловой энер
гии, выделяемой отходами в течение длительного периода времени.
В настоящее время нельзя также для захоронения рекомендовать какую-либо опре
деленную геологическую формацию, например, каменную соль. Поэтому в программе
на равных правах рассматриваются каменная соль, глины, скальные породы (порфирит,
гранит, базальт и др.).
Второй особенностью программы является комплексность исследований, ос
нованных на применении современного физико-математического аппарата, контроль
но-измерительной аппаратуры и моделирования. Программой предусматривается ши
рокое использование математического моделирования с применением ЭВМ. Задачи о
напряженно-деформированном состоянии массива и обделок планируется решать с
помощью моделей, выполненных из оптически чувствительных и эквивалентных матери
алов, что позволит оценить одновременное действие на массив веса налегающих пород,
внутреннего давления и температуры, развиваемых отходами, размещенными в герме
тичном подземном хранилище. Экспериментальные исследования будут проводиться
с применением оптико-поляризационных методов и стендов объемного нагружения.
Большая роль в программе отводится опытам, проводимым в реальном горном массиве
с имитаторами, представленными электронагревателями, отработанными тепловыделя
ющими элементами и другими источниками активности. В частности, предполагается
изучить эффекты, сопровождающие интенсивный прогрев сферической емкости, соз
данной методом размыва, в каменной соли на глубине 100 м. Намечено провести опыт
в буровых скважинах, пробуренных в глине.
Значительное внимание в программе уделяется опробованию в условиях, близких
к натурным, и совершенствованию конструкционных материалов (цемент, металл, поли
меры и др.). Предполагается использование в исследованиях агрессивных сред и ради
ационного облучения.
Придавая большое значение вопросам оценки пригодности геологических участ
ков для захоронения отходов, в программе предусмотрено дальнейшее развитие при
меняемых в геологоразведке методов, включая электро- и сейсмозондирование, грави
метрию, нейтронный каротаж, ультразвуковое просвечивание образцов, комплексное
испытание кернового материала, опытные нагнетания в скважины жидкости и газа и др.
Предусматриваются также работы по оценке возможности проявления в районе распо
ложения намеченных под захоронение отходов участков тектонических процессов.
Самостоятельным в программе является раздел, посвященный конструированию
систем и средств контроля за процессами, происходящими в хранилище и окружающем
горном массиве, а также охлаждающих систем. При этом особое внимание обращено
на долговечность средств и систем и их замену в процессе эксплуатации хранилищ.
38 КЕДРОВСКИЙ и др.
Вполне естественно, что каждую научно-техническую разработку намечено прово
дить применительно к потенциально осуществимым схемам (вариантам) захоронения
отходов. Ниже приводится краткое описание некоторых из этих схем.
1. Захоронение отвержденных отходов в шахтных стволах
Шахтный ствол диаметром 4-6 м приходится на глубину 400-800 м с гидроизоля
цией и тюбинговой крепью. Небольшое количество воды, просачивающийся через крепь,
собирается в зумпфе и откачивается на поверхность погружными насосами, опущенны
ми через трубные каналы. Отвержденные отходы укладываются послойно в ствол с за
сыпкой или перемешиванием с кварцевым песком. После заполнения ствола до отмет
ки, расположенной на 150-250 м ниже поверхности земли, проводится ’’сушка” загру
женной смеси и прогрев крепи до температуры более 100° С с отводом парогазовой сме
си. Затем над верхним слоем отходов возводится перемычка с контрольными трубны
ми каналами, снабженными запорной арматурой. Этим начинается период консервации
хранилища. Опасность выщелачивания активности подземными малоинтенсивными во
дами исключается наличием теплового барьера. Возможны и другие варианты конст
рукций хранилищ в шахтных стволах.
2. Захоронение отвержденных отходов в буровых скважинах
Вариант 2.1. Для размещения пеналов с отвержденными отходами в толще слабо про
ницаемых пород бурится серия скважин. Пеналы с отходами в зависимости от величи
ны тепловыделения размещаются либо непрерывно друг за другом, либо последователь
но чередуются с пробками. Заполнение скважин пеналами производится до глубины,
обеспечивающей изоляцию отходов от горизонта водообмена. Оставшаяся свободная
часть скважины цементируется до устья.
Вариант 2.2. В том случае, если возможна организация полигона захоронения в мало
населенной местности, в которой и в перспективе не предвидится строительства жилищ
ных и промышленных комплексов при наличии соответствующих геологических усло
вий, возможно устройство хранилища небольшого заглубления. В этом случае пеналы
с отходами размещаются в буровых скважинах, пробуренных в толще мощных глинис
тых отложений на глубину до 40 м. После размещения пеналов верхняя часть скважи
ны цементируется. Мощность толщи глин должна быть таковой, чтобы имелась доста
точная изоляция ее от зоны водообмена.
Недостатком этого варианта является малая надежность при нарушениях поверх
ностного слоя земной поверхности от стихийных бедствий или иных экскавационных
воздействий.
3. Захоронение отвержденных отходов в подземные емкости
Для захоронения больших объемов отвержденных отходов могут быть исполь
зованы специально сооружаемые подземные емкости. Независимо от типа горных
пород и способа сооружения таких емкостей отвержденные отходы в пеналах или кон
тейнерах подаются в камеру захоронения через шахтный ствол или буровую скважи
ну, которые снабжены соответствующими шлюзовыми устройствами. После заполнения
IAEA-SM-243/1 IS 39
емкости планируемым количеством отходов ствол или шахта цементируются на всю
высоту.
Вследствие накопления большой массы отходов происходит выделение тепла,
способного расплавить окружающую породу, которая вместе с расплавленными отхо
дами образует высокотемпературный очаг. Из-за разности объемных весов расплава
и окружающих горных пород в течение всего времени выделения тепловой энергии
будет происходить его постепенное опускание. Существенным элементом такой
технологической схемы является необходимость выбора такого участка, в пределах
которого за все время существования очага исключалась бы возможность контакта
содержащихся в нем продуктов с зонами, по которым возможен их вынос в биосферу.
Сооружение самих хранилищ возможно несколькими методами.
4. Захоронение отвержденных отходов в подземных горных выработках
Одним из методов захоронения пеналов или иных контейнеров является разме
щение их в шахтных выработках. Доставленные к месту захоронения в шахте отходы
могут размещаться либо в специальных камерах, либо в буровых скважинах. После
размещения отходов в скважинах, камерах и подходных выработках устанавливаются
герметизирующие пробки, перемычки и тому подобные конструкции.
Схемы организации хранилищ отходов могут осуществляться по двум вариантам.
Вариант 4.1. Использование для сооружаемого хранилища отработанных рудников
или шахт с сооружением в них специальных камер.
Вариант 4.2. Сооружение специальных шахт с системой выработок и камер для захо
ронения.
К настоящему времени в соответствии с комплексной программой рассмотрено
несколько вариантов захоронения отходов в шахтные стволы и скважины; укрупненно
решены задачи о параметрах температурных полей в массиве при различной характерис
тике источников тепла, представленных высокоактивными отходами; оценена картина
распределения в массиве температурных напряжений; проведены исследования в натур
ных и лабораторных условиях некоторых горных пород, включая каменную соль, пор-
фириты, гранит, песчаник и др. ; изучены геологическое строение и гидрогеология на не
которых участках, представляющих интерес для захоронения отходов; изучено поведе
ние некоторых конструкционных материалов, находящихся в условиях длительного тем
пературного и радиационного воздействий:
Научно-исследовательские и опытно-конструкторские работы по программе про
должаются.
DISCUSSION
G.A.M. WEBB: With reference to the fundamental criteria, I would like to
question one which concerned the ensuring of isolation for an unlimited period.
At times well into the future when some radionuclides are still present, the uncer
tainty of predictions will increase and it will be very difficult to demonstrate with
40 КЕДРОВСКИЙ и др.
confidence that these radionuclides will never escape. Such a criterion may there
fore be unwise. Would it not be better to have a more limited objective, such as
to limit the amount released to an acceptable level? Fulfilment of this objective
could be satisfactorily assured by predictive modelling assessments.
M.K. PIMENOV: In speaking about the need for isolation of wastes for an
unlimited period of time, we mean the very long period of storage which is needed
for the activity to decrease to a safe level. According to calculations, this period
may be as long as hundreds of thousands and even millions of years.
Your idea about developing criteria permitting release of small amounts of
activity (bearing in mind permissible levels) is worth considering.
J.J.K. DAEMEN: Dr. Pimenov mentioned the introduction of plugs between
containers in deep shafts or boreholes and also the cementing of shafts. Would
he please specify what materials are to be used for the plugs and what studies have
been performed in this connection.
M.K. PIMENOV: For plugs it is intended to use quartz sand, gravel, tamping
cement of the “hot-well” type and so on. Investigations of such materials are now
under way.
P.A. WITHERSPOON: Have you set a maximum temperature which the
repository in crystalline rock will not be allowed to exceed?
M.K. PIMENOV: No, we have not set a temperature ceiling. In the USSR
we are considering a number of options for the disposal of high-level solidified
waste in geologic formations, including some where the maximum temperature is
that at which the waste and the surrounding rock melt.
P.A. WITHERSPOON: Have you developed a method of predicting the
movement of radioactive waste through the fracture systems in hard crystalline
rocks?
M.K. PIMENOV: To predict the movement of radioactive waste in fissured
rocks we use theoretical calculations and analytical dependences characterizing
the hydrodynamics of this process. These are well-developed techniques used
in the oil and gas industry.
R.G. CHARLWOOD: Could you please explain the particular sources of
the high pressures which you stated are expected to exist in the high-level waste
repository area? Are these due to processes other than thermal expansion of
solids?
M.K. PIMENOV: These high pressures are due precisely to the thermal
expansion of rocks in the repository.
IAEA-SM-243/30
NATIONAL POLICY FOR UNDERGROUND
DISPOSAL OF RADIOACTIVE WASTES
IN THE UNITED KINGDOM
F.S. FEATES
Radioactive Waste Management (P) Division,
Department of the Environment,
London, United Kingdom
Abstract
NATIONAL POLICY FOR UNDERGROUND DISPOSAL OF RADIOACTIVE WASTES
IN THE UNITED KINGDOM.
Legislation controlling the disposal of radioactive wastes within the United Kingdom
is summarized. The Department of the Environment now has responsibility for developing
the national policy of radioactive waste management and is also the body responsible for
authorizing disposal. Present practice involves the use of shallow land burial sites of various
kinds for wastes of low activity. Research is being undertaken to determine the suitability
of UK geological formations to contain long-lived wastes of higher activity in a safe and
environmentally acceptable manner. United Kingdom policy requires a full assessment of all
options for the disposal of wastes to be undertaken so that the optimum solution appropriate
to national needs can be selected. A detailed systems analysis of wastes and their disposal
requirements (including use of the oceans as disposal media) is therefore being initiated with
the objective of completing the assessment within the next decade.
1. LEGISLATION
The main legislation governing the accumulation and disposal of radioactive
wastes in the United Kingdom is the Radioactive Substances Act (1960). The
provisions of the Act relating to accumulation of radioactive wastes apply to
premises such as hospitals, research institutes and industrial premises but not to
the UK Atomic Energy Authority (UKAEA) premises, licensed nuclear sites
or Crown properties. The provisions of the Act relating to disposal of radioactive
wastes apply to all premises other than Crown property, including the UKAEA
and licensed nuclear sites. The Act prohibits both the accumulation and
disposal at or from the relevant premises unless authorized by the appropriate
Ministers.
Responsibility for the administration of the Radioactive Substances Act 1960
lies with the Secretary of State for the Environment in England, his colleagues
in Scotland and Wales and, in Northern Ireland, the Department of the Environment.
41
42 FEATES
These Ministers alone are responsible for authorizing the accumulation and
disposal of radioactive wastes on and from any premises other than the UKAEA
and licensed nuclear sites. In England authorizations for disposals on the last
two categories of site are jointly issued by the Environment Secretary and the
Minister for Agriculture, Fisheries and Food. In Scotland and Wales the
Secretaries of State act alone since they represent both the agricultural and
environmental interests of their territories.
Any authorization issued under the Act may be granted subject to
any limitations or conditions the Minister thinks fit; and it may be varied
or revoked. Before issuing authorizations for waste disposals from UKAEA
and licensed nuclear sites Ministers are, it should be noted, obliged to consult
whichever public or local authorities they consider appropriate, e.g. regional
water authorities, district and county councils. Consultation is also obligatory
in cases where disposals are made to local authority landfill sites and special
precautions are required. The Act also empowers the Environment Secretaries
to provide or to arrange the provision of radioactive waste disposal facilities if
they consider that adequate facilities are not available.
The Radioactive Substances Act (1960) is at present being closely studied
to establish whether, in view of changes brought by the passage of time, it
remains adequate for its purpose. In this context it is worth noting that if,
in future, a decision is taken to construct an underground land repository for
waste,, new legislation, or at least additional legislation, may be required to
regulate such a repository. The Department of the Environment will obviously
take a leading part in promoting any such legislation because of the clear
environmental implications of such a decision.
2. PRESENT PRACTICE IN THE DISPOSAL OF RADIOACTIVE WASTES
FROM ALL SOURCES, INCLUDING NUCLEAR POWER GENERATION
Authorizations permitting the disposal to land of solid radioactive waste of
very low activity with ordinary refuse are granted under the Radioactive
Substances Act. No control beyond the point at which the disposer puts the
waste in his dustbin is required. Wastes of somewhat higher activity can still
be disposed of to a landfill site provided certain precautions are taken. Authorizations
for such disposals specify the landfill which is chosen after consideration of its
management, its expected life, the probable subsequent use of land, whether it
is liable to catch fire, drainage, local water supplies and any other special features.
Authorizations are exceptionally granted for the disposal of solid radioactive
waste by burial on the site at which it arises. Amongst the additional criteria
for disposals of this kind is some assurance of continuity of ownership of the
site bearing in mind the activity and half-life of the waste.
IAEA-SM-243/30 43
Solid wastes of higher activity than those described above are disposed of
through the National Disposal Service (NDS) which is operated by the United
Kingdom Atomic Energy Authority and British Nuclear Fuels Limited (BNFL)
acting as agents for the Secretary of State for the Environment. At present the
only site in the UK for shallow land burial, in trenches, of waste arising through
the NDS is at Drigg on the coast of Cumbria, in northwest England. Certain
BNFL and UKAEA sites dispose of limited quantities of waste by burial within
the site boundary.
Where disposal is made to landfill sites neither the disposer of the waste nor
the operator of the site is required to monitor the surrounding environment.
A full monitoring programme is, however, carried out by BNFL at Drigg. Any
necessary monitoring of sites other than Drigg is carried out by the Radiochemical
Inspectorate of the Department of the Environment in England and Wales, the
Industrial Pollution Inspectorate in Scotland, and by the Nuclear Sites Inspectorate
of the Ministry of Agriculture, Fisheries and Food. Wastes unsuitable for shallow
land burial are not at present authorized for underground disposal in the UK.
They are either packaged and disposed of at sea (under licence from the Minister
of Agriculture, Fisheries and Food in England, or the Secretaries of State for
Scotland or Wales, depending on the port of loading, but in all cases under the
provisions of the 1974 Dumping at Sea Act); or they are stored in silos; or,
in the case of highly active wastes from reprocessing, in special high-integrity
stainless steel tanks. All these disposal and storage practices are subject to
Government approval, careful inspection and, in the case of sea disposal, international
surveillance and controls.
3. ADMINISTRATIVE DEVELOPMENTS
The Standing Royal Commission on Environmental Pollution (RCEP)
published its Sixth Report “Nuclear Power and the Environment” in September
1976. It recommended that the responsibility for developing the strategy to
deal with radioactive wastes should lie with the Government Departments
concerned with the protection of the environment rather than with the Depart
ment responsible for developing and promoting nuclear power. In its response
to this recommendation, the Government stated that the Secretary of State
for the Environment would in future be responsible, together with the
Secretaries of State for Scotland and Wales, for nuclear waste management policy.
The main elements in this new responsibility included securing the disposal of
wastes accumulated and arising at nuclear sites and ensuring that adequate
research and development is undertaken on methods of disposal.
44 FEATES
The Royal Commission also recommended that there should be no commit
ment to a large-scale nuclear programme in the UK until it had been demonstrated
beyond reasonable doubt that a method existed to ensure the safe containment
of long-lived highly radioactive waste for the indefinite future. In its response
the Government pointed out that it shared the confidence expressed by the
Royal Commission that a solution to this problem would, in fact, be found.
The Government went further to say that the question of the safe containment
of highly active wastes was bound to be the dominant factor in any process
preceding decisions about further large-scale nuclear programmes.
Another important recommendation made by the Royal Commission, and
accepted by the Government, was that an independent committee should be
set up to advise Ministers on the broad policy issues affecting radioactive, waste
management. In May 1978 the Radioactive Waste Management Advisory
Committee was.therefore established under the chairmanship of Sir Denys
Wilkinson, FRS. Its terms of reference are:
“To advise the Secretaries of State for the Environment, Scotland and
Wales on major issues relating to the development and implementation
of an overall policy for the management of civil radioactive wastes;
including the waste management implications of nuclear policy, of the
design of nuclear systems and of research and development; and the
environmental aspects of the handling and treatment of wastes.”
The Government is at present considering whether the Secretaries of State
for the Environment, Scotland and Wales will need, in addition to those powers
already available under existing legislation, further statutory powers to help
them in carrying out their new responsibilities for radioactive waste management
strategy. Since the present system of controls was developed nearly 20 years
ago, and important changes have occurred since then, the Department of the
Environment appointed in March 1976 (before the Royal Commission’s
Sixth Report was pubUshed) a group of experts to review the adequacy of present
arrangements. Their report will shortly be considered by the Radioactive Waste
Management Advisory Committee.
Since this review of present arrangements was still in progress when the
Government responded to the Royal Commission’s Sixth Report, the Govern
ment considered that it would be premature to decide the merits of the Royal
Commission’s recommendation that a Nuclear Waste Disposal Corporation
should be set up to develop and manage existing and new radioactive waste
disposal facilities. The Government therefore undertook to reconsider this
proposal in due course in the light of the review’s conclusions and the advice of
the Radioactive Waste Management Advisory Committee.
IAEA-SM-243/30 45
Existing shallow land burial procedures remain suitable for disposal of very
low level solid radioactive wastes and no specific research is required into
management or disposal of these wastes. For those wastes which require a
special landfill site operated under strict surveillance, the Drigg site is the only
one available on a commercial basis in the UK, so there is a need to develop one
more such site to provide a back-up capability and to minimize transport.
Limited geological and hydrogeological research will be required to identify a
suitable site. The main wastes in the UK for which land disposal routes have
not been fully evaluated are typical of those arising from a nuclear power
programme with reprocessing of spent fuel. These are high-level, heat-generating
liquid waste, solid fuel element cladding and related waste from reprocessing
operations and power reactor wastes such as spent ion-exchange resins, sludges
and redundant irradiated items. In addition, the future will see arisings of waste
from the decommissioning of reactors and other equipment such as reprocessing
plant. The difficulty of disposal is greatest in the case of heat-generating wastes.
Their total bulk is relatively small and work is proceeding on their vitrification,
which will aid the eventual disposal in a safe manner. Research is also being
undertaken to establish disposal routes for the very much larger quantities of
the other wastes.
It is too early to say with certainty what land-based facilities will be required
in the United Kingdom, since sea disposal is at present adoptee! for the reactor
wastes which comply with the requirements of the London Convention on
Dumping of Wastes at Sea. Also, the UK is participating in an international
programme of research into the feasibility of disposal of solidified heat-generating
wastes in the deep ocean. However, it is probable that two further types of
land-disposal facilities will be necessary in addition to the existing shallow land
burial facility. These will provide respectively for the disposal of heat-generating
and high and intermediate level non-heat-generating solid wastes. Such facilities
might or might not be in the same location and, indeed, it is the overall
objective of the UK underground disposal research and development programme
to establish if one or more repositories can be constructed underground to
provide for the safe disposal of all wastes, including those from decommissioning
reactors and other nuclear plant.
Guidelines have already been drawn up by the Institute of Geological
Sciences for the selection of potentially suitable geological formations for the
disposal of heat-generating wastes below land. Emplacement in crystalline rock
(like granite), in argillaceous formations (like clay) or in evaporite deposits
(like salt) holds most promise. In all, about 18% of the UK land area appears
to comply with the requirements of the above-mentioned guidelines. Research
into disposal of heat-generating wastes started in the UK over two years ago
4. TECHNICAL DEVELOPMENTS
46 FEATES
and consent is to be sought for test-drilling in about 20 locations. These cover
different rock types and structural settings. So far, consent has been granted for
one location in crystalline rock in the north of Scotland, where work is proceeding.
It is hoped to select two or three of the most promising sites for detailed
appraisal by 1983. This appraisal will involve consideration of the suitability
of each site for the complete range of wastes requiring disposal and of the
appropriate depth for safe disposal of these various wastes. Heat dissipation
studies are also most important for estimating the configuration and size of a
repository for heat-generating waste and for determining the requirements of
pre-disposal for storage of waste. Preliminary in-situ studies of thermal effects
in granite are already well advanced at an experimental facility in Cornwall.
Amongst the options which will be considered will be the mixing of heat-generating
and non-heat-generating wastes within the repository and also segregation of
the various types of waste at different levels.
By 1984 it is expected that the initial appraisal of the potential of ocean
disposal for these wastes will also have been completed so that all disposal
options may then be assessed.
After 1984 a further five years of detailed field studies will be required
before disposal of even non-heat-generating wastes can be considered. Probably
ten years’ work will be required to determine if the site can be used for disposal
of heat-generating wastes. Disposal of non-heat-generating waste may therefore
be possible by 1990 with pilot disposal of the heat-generating waste following
five years later. Operational facilities for heat-generating waste are unlikely to
be required until after 2005, allowing a further ten years for detailed development
work and assessment.
The geological and radiological assessment studies which will form a major
part of the required research programme are described in other papers at this
Symposium. In addition to these a wide range of supporting studies will be
necessary before the safety of underground disposal can be established. For
example, the assessment of migration of radionuclides through a geological
formation requires, in addition to hydrological information, an extensive knowledge
of radionuclide/groundwater chemistry and of the mechanisms of sorption on rock.
At this stage it is not clear that it will be possible to show convincingly that
geological formations will be adequate long-term barriers against dispersion of
dissolved radionuclides back to the environment. Therefore detailed studies
are under way on man-made barriers which can provide further isolation of the
waste. These involve study of waste-packaging methods, construction of adsorber
beds and, in the future perhaps, even study of improved matrices for waste
encapsulation. (It is already planned to employ man-made barriers with sufficient
integrity to contain the waste for at least 500 years by which time the highly
radioactive fission products will have decayed.) Engineering and underground
construction studies are required to develop reference designs and, at a later
IAEA-SM-243/30 47
stage, site-specifíc designs for repositories. These will provide for disposal of
heat-generating and non-heat-generating wastes from reprocessing operations,
power reactor wastes and material from reactor decommissioning. As an insurance
against cessation of reprocessing they will also provide for the disposal of
unreprocessed spent fuel elements as well as, or instead of, reprocessing wastes.
These designs will be used for assessment of construction costs and of the safety
of operating procedures. Island, coastal and inland sites will be evaluated and,
since vitrified wastes may require storage for several decades prior to disposal
so that the heat-generating capacity is reduced, a combined store/repository
concept will be investigated. A closely related set of engineering studies will
consider the influence of nuclear heating on repository integrity. These studies
involve prediction of the temperature field within and around a repository in a
geological formation, of the consequent thermal stresses in the surrounding rock,
of the stress-induced fracturing of the rock and of the resulting effects on
groundwater flow and waste migration. The programme will also include studies
of backfilling and sealing of repositories and of corrosion due to interactions
between rock fluids and waste-containment materials.
In conjunction with the studies described above a systems study of waste
management strategy will be carried out. The objective of this is to define a
national strategy for the safe, cost-effective disposal of all categories of waste.
This will require optimization of all technically feasible disposal strategies in
accordance with the principles of ICRP Publication 26. This study will identify
interaction between technical options and choices within any given strategy for
a.particular waste category, such as choice of waste form, duration of interim
storage or repository host medium, and interactions between waste categories
resulting from treatment options such as incineration, acid-digestion, evaporation
or decontamination. Until it is possible to define safe disposal strategies,
government will be able to ensure, by this means, that adequate research and
development is being undertaken. In addition, a most important product of
this study will be the establishment of waste form criteria which' will allow
waste producers to direct their waste conditioning development efforts appropriately.
A growing proportion of this investigation forms part of the CEC Indirect
Action Programme of research into the disposal of waste in geological formations.
The whole programme is co-ordinated with research in other countries through
relevant OECD/NEA and IAEA review committees.
DISCUSSION
H.O. BÓHM: You mentioned that there are plans for employing man-made
barriers of sufficient integrity to contain the waste for at least 500 years. Are
you thinking of thick-walled containers and specific container materials? What
other specific types of barrier are you considering?
48 FEATES
F.S. FEATES: Metal or ceramic canisters are being considered as well as
chemical and physical man-made barriers like bentonite.
G. ROCHLIN: Your presentation seemed to express a general attitude
that there was not much urgency for acquiring a repository, largely because
surface or near-surface storage of high-level waste is a satisfactory and acceptable
option, at least for the remaining years of this century. Will interim storage of
high-level wastes be limited solely to those of British origin, or will this service
also be offered to your overseas reprocessing customers? Will storage be in
liquid or solid form?
F.S. FEATES: Wastes will initially be stored in liquid form. At this stage
all reprocessed waste will be retained in the United Kingdom. After vitrification
waste may be returned to the country of origin and all contracts for reprocessing
wastes originating from outside the United Kingdom will include provision
for their return.
H. KRAUSE: You stated that very-low-level wastes were disposed of by
shallow land burial and that wastes of a somewhat higher level were dumped into
the sea. Have you established your criteria and upper limits for accepting waste
for land burial in terms of maximum specific activity or in terms of a potential
hazard over a certain period of time (e.g. 50 years, 100 years etc.)?
F.S. FEATES: Wastes for land burial are authorized on a site-specific
basis. The geology of the sites and the nature of ground and surface water
movements are assessed before the levels authorized are decided. As for landfill
sites operated by local authorities, only short-lived isotopes may normally be
disposed of there.
J.J.K. DAEMEN: You mentioned backfilling studies as a critical aspect of
the effort to eliminate what is otherwise possibly the shortest leakage path.
Could you give details of your backfilling studies?
F.S. FEATES: The backfilling programme is now being formulated. No
work has yet been completed.
IAEA-SM-243/158
UNDERGROUND DISPOSAL OF
RADIOACTIVE WASTES IN INDIA
Past experience and future planning
K.T. THOMAS, N.S. SUNDER RAJAN, K. BALU,
M.P.S. RAMANI
Bhabha Atomic Research Centre,
Trombay, Bombay, India
Abstract
UNDERGROUND DISPOSAL OF RADIOACTIVE WASTES IN INDIA: PAST EXPERIENCE
AND FUTURE PLANNING.
Engineered facilities for storage and disposal of low- and intermediate-level solid wastes
have been in operation at different locations in India for a number of years. Criteria for
selection of sites, design and experience in operation of these facilities are reviewed. For
disposal of solidified high-level wastes, the work carried out so far has been limited to identi
fication of a few candidate sites for location of a repository. This work has been summarized
in the paper, which outlines the future programme of work in this field.
1. INTRODUCTION
In the Indian nuclear programme, all solid wastes generated in the past have
been mostly low- and intermediate-level beta-gamma wastes and alpha-contaminated
wastes. The solid wastes produced vary from paper, glass, plastic and rubber items
to contaminated equipment and piping, fuel cladding hulls and other non
combustible materials. In addition there are other wastes such as spent resins,
chemical sludges, and evaporator concentrates which have to be solidified. The
wastes are effectively segregated at source. Combustible primary solid wastes are
double packaged in polythene and pvc bags and are collected in 200-litre carbon
steel drums. The spent high-efficiency particulate filters are collected in pvc bags.
The process equipment and piping are decontaminated and painted prior to
collection.
For effecting volume reduction, a 15-ton hydraulic baler is used. Incineration
is used to a limited extent at one of the sites for low-level beta-gamma wastes.
For alpha-contaminated wastes so far no conditioning or volume-reduction methods
have been employed. A pilot incinerator for this purpose is under development.
The evaporator concentrates and chemical sludges are incorporated in concrete/
bituminous matrices. ‘Solcofloc’ or other drying agents have also been used for
fixation of filter sludges.
49
TABLE I. SALIENT FEATURES OF WASTE MANAGEMENT SITES IN INDIA
Soü properties
Site General geology Sub-surface features HydrologyType PH CECa
meq/g
Kd
ml/g
1. Site 1 — Rese.arch
Centre I
Igneous extrusive base
rocks; mainly basalt with
Amygdaloidal tuff etc.
at places
Up to 4.5 m — soil
4.5—6.5 m — weathered
rock. Below 6.5 m -
bluish gray basalt
Groundwater confined to
soil and weathered zone
mostly during the
monsoon season
(i) Groundwater velocity
in soil: 2 cm/d
(ii) Permeability in
weathered rock:
0.9 m/d
Süty
loam
clay
7.9-8.2 45-50
2. Site 2 — Power
Station I
3. Site 3 — Ore
Beneficiation
Facility I
Igneous crystalline and
metamorphic rocks
Up to 0.9 m — soil
0.9—1.5 m — soft
laterite; below
1.5 m - hard laterite
Süty
loam
clay
5.5-6.9 25 15-30
4. Site 4 — Power
Station II
Same as above Up to 5.00 m —
soil 5—8 m —
weathered rock below
8.0 m — chamockite
Groundwater velocity —
15 cm/d
Sandy
loam
clay
About 7 5-17
TH
OM
AS
et al.
Site General geology Sub-surface features HydrologyType
Soil properties
pH CECa
meq/g
Kd
ml/g
5. Site-5 — Fuel
Fabrication
Facility
Igneous crystalline and
metamorphic rocks
Up to 1 m — soil and
weathered rock.
Below 1 m — hard and
compact pink-gray
granite-gneiss
Groundwater in
limited quantities at
6—8 m depth
Sandy
loam
6.6-7.8 18-20 19-64
6. Site-6 — Ore
Beneficiation
Facility II
Same as above Up to 20 m - soil
Below 20 m —
mica-schist
Water table at 4 m SUty
clay
and
loam
6.6—8.0 10-48 16-60
7. Site-7 — Power
Station III
Sedimentary formations -
unconsolidated argillaceous
rocks
No soil capping;
hard and compact
quartzitic sandstone
with two sets of joints
and bedding planes
No groundwater
8. Site-8 — Power
Station IV
Same as above Up to 6.5 m — soil
Below 6.5 m —
micaceous sand
Groundwater velocity
1.06 cm/d
Silty
loam
7.5-7.9 12-23
a Cation exchange capacity.
IAEA
-SM
-243/1
58
52 THOMAS et al.
High-level radioactive wastes generated from fuel reprocessing plants will be
solidified by incorporation in suitable vitreous matrices. The soUdified wastes
will be given interim storage on site.
2. SITE CRITERIA FOR NEAR-SURFACE STORAGE AND DISPOSAL
FACILITIES
Nuclear facilities, such as power reactors, uranium and thorium mining and
milling facilities, fuel fabrication facilities, research centres, etc., are located in
different parts of India. Due to the vastness of the country and problems of
logistics, considerations of both economy and safety have warranted location of
individual on-site storage facilities for low- and intermediate-level solid wastes.
Varying geological, hydrological, climatic and other environmental conditions
are found at these sites. The sites can be grouped geologically as follows:
(1 ) Igneous rock formations: Basaltic at the Tarapur Atomic Power Station
and Bhabha Atomic Research Centre, which are coastal sites.
(2) Igneous, crystalline and metamorphosed rocks: Granitic at the Madras
Atomic Power Project site, the Nuclear Fuel Complex site at Hyderabad
and the Uranium Corporation of India Ltd. site in Bihar. The Madras site
is coastal and the other two sites are inland.
(3) Sedimentary Rocks:
(a) Argillaceous and unconsolidated formations such as clay, silt and sand
for the Narora Atomic Power Project site in Uttar Pradesh, which is in
a fertile land-locked site.
(b) Consolidated sandstone at Rajasthan Atomic Power Project site, which
is an inland site and facing a lake.
To evaluate the suitability of sites for on-site storage of radioactive wastes,
extensive site investigations were carried out. Groundwater movement studies
using active and inactive tracers and pumping tests were conducted. Detailed
laboratory studies of the soil and rock supplement the field data.
Some of the results of these investigations are presented in Table I.
It has been observed that groundwater fluctuates considerably at almost all
the sites, depending on the intensity of rainfall. A major portion of the country
receives moderate to very heavy rainfall during the June-September period. At
places like Tarapur, receiving around 200-250 cm of rainfall in this period, the
underground level reaches to within a few centimetres of the surface, but recedes
fast after the end of the monsoon to depths of more than 3 metres and even dries
up, depending on the subsurface geological condition. At coastal sites the ground
water flow is towards the sea and at the Narora Power Project site, it was observed
to flow towards the Ganga canal located in the north. At the Rajasthan Power
Station site, the bedrock hard, bedded and jointed, the rainwater percolates through
the fissures and channels into the large lake near the site.
IAEA-SM-243/158 53
The above investigations have led to the design of engineered containments
to suit the specific site conditions. For instance, solid storage is not permitted at
Narora and at the Rajasthán site the engineered containments for waste storage
are located above surface. Some of the criteria for the design of these near-surface
storage facilities are given below:
(1) Direct disposal into the ground is not permitted.
(2) Liquid wastes are not stored in these facilities. Liquids, sludges and evaporator
concentrates are converted into suitable solid forms by incorporation in
bituminous or concrete matrices.
(3) Leakages from the storage and disposal facilities are to be as low as is
practicable.
(4) A number of barriers are incorporated in the design of the facility to inhibit
migration of radionuclides to the environment.
The waste is retrievable in all these facilities though a definite need for
retrieval is not indicated at this stage.
3. OPERATING FACILITIES
The earliest near-surface storage and disposal facility constructed was the
Radioactive Solid Storage site at the Bhabha Atomic Research Centre at Trombay.
This has been in operation for around fifteen years. The near-surface disposal
facilities located at Tarapur receive radioactive wastes from the 400 MW power
station and also from the fuel reprocessing plant located there. The facilities
have been in operation for the past seven years. Table II presents the solid-waste
generation in Trombay and Tarapur sites in the last three years. As can be seen,
at Trombay, the total generation during 1978 was around 500 m3 with about
20% of it alpha-contaminated. These quantities do not include cladding hulls
and degraded solvent from fuel reprocessing plants. A little more than 75% of
the alpha-contaminated wastes consist of combustible material such as paper,
cotton waste and gloves. The rest of it is mainly glassware, small metallic com
ponents, metal containers and filters. Of the other wastes about 25% are drums
containing chemical sludges and spent resin incorporated in concrete/bituminous
matrices and about 5% are spent filters. The rest is miscellaneous.
At present all the primary solid wastes generated from the various facilities
are stored in near-surface concrete vaults and are retrievable. The concrete vaults
are 1 0 m X 3 m X 4 m deep and are internally surface-finished with bitumastic
paints. Different vaults are used for storage of different types of wastes. Figure 1
shows a trench at Tarapur site before closure. The top of the trench is closed by
prefabricated concrete slabs of suitable thickness. A special coal-tar-based epoxy
treatment is given over the top finished plaster, to serve as a waterproof lining.
t-Л-Рь
TABLE II. GENERATION OF RADIOACTIVE SOLID WASTES AT BARC TROMBAY AND TARAPUR SITES
Beta-gamma wastes Alpha-contaminated wastes Total
Year (mS) (m3) (m3)
Trombay Tarapur Trombay Tarapur Trombay Tarapur
1976 650 500 100 50 750 550
1977 390 452 100 38 490 490
1978 426 629 84 68 510 697
TH
OM
AS
et aL
IAEA-SM-243/158 55
FIG.l. View o f the reinforced cement concrete trenches at the Tarapur site.
For higher levels of solid radioactive wastes showing surface doses greater
than 100 R/h and for significant transuranic solid wastes, specially engineered
“tile holes” are being used. These are a series of underground circular vaults of
710 mm inside diameter and 4 metres in depth. Spun concrete pipes are used for
the primary construction and these are lined with 6 mm thick steel plates which
are again sandwiched between layers of spun concrete. These steel-lined concrete
pipes are embedded in a pad of reinforced concrete at the bottom with a steel
plate welded to the bottom of each pipe, thus providing in effect a steel box
embedded in spun concrete. Additional protection of a waterproof lining is given
to all the external surfaces of each such tile hole.
The higher-level solid wastes such as fuel components, zircaloy hulls, spent
resins, and filter cartridges are received in special steel drums and are placed in the
tile holes remotely. Figure 2 presents a view of the tile holes in use at the Tarapur
site. The integrity and leaktightness of such a facility has been extensively tested
for several years. The cost of such storage is estimated to be about US $4000/m3
of wastes stored. Storage of cladding hulls is considered only as an interim approach.
56 THOMAS et al.
FIG.2. A view o f the ‘‘tile holes”at the Tarapur site.
No decision has been taken on their ultimate dispensation, though at present
development work is underway for compaction of these wastes.
The third major nearrsurface storage and disposal site in operation is at
Rajasthan near the power station site. As mentioned earlier, the geology of the
site is unsuitable for location of an underground disposal site. The bedrock,
sandstone, is highly fissured and jointed. Due to the lack of overburden at this
site, rainwater directly channels to the large surface water body at the site through
the bedrock. As such the storage and disposal facilities such as trenches are located
overground, protected by suitably designed soil embankments. Higher levels of
radioactive wastes generated at this site are transported for off-site disposal.
Similarly, wastes generated at Narora site will be compacted, packed and shipped
off-site for disposal, since the site is fertile, landlocked and the underground
aquifer is extensive and connected to nearby canals and rivers.
At each of the above sites, a number of post-operational monitoring wells
are located to facilitate routine surveillance of the integrity of the facilities. The
experience of the past fifteen years has built up confidence that such facilities
will meet the stringent safety standards set by the health and safety authorities.
IAEA-SM-243/158 57
4, DISPOSAL OF HIGH-LEVEL RADIOACTIVE WASTES
The problem of handling and disposal of high-level wastes is at present
receiving special attention. Though the rate of generation of such wastes is rather
small at present, it is likely to increase substantially with our commitment to
reprocess spent fuel and recycle fissile material. The present policy of manage
ment of high-level wastes is based on immobilization of the wastes in suitable
vitreous matrices, provision of cooled storage for solidified wastes under sur
veillance for an interim period and finally disposal in deep geological formations.
The type of matrix to be used for immobilization and the characteristics desired
of the end-product depend on the rock medium in which disposal is effected.
At present the first plant for immobilization of high-level radioactive wastes is
at an advanced stage of construction at Tarapur. The solidified wastes are to be
stored on-site for a period of about 20 years. The storage is based on natural
convective air cooling.
For disposal of the solidified wastes, the work under way mainly pertains
to identification of a few candidate sites suitable for the location of a geological
respository.
4.1 . General features
India is a vast country with contrasting geographical and geological features.
From north to south these features include the Himalayas, Indo-Gangetic alluvium,
Thar desert, the upper Cretaceous volcanic spread (Deccan traps), the Precambrian
sedimentary formations including the Cuddappahs and Vindhyans, the coastal
plains with Tertiary and Recent sediments, and the Precambrian crystalline
complex of the peninsula.
On the basis of the criteria required for a candidate site for a repository,
many of the above areas appear to be ruled out on the basis of data available.
The Himalayas lie in the orogenic belt with a very high seismic risk. Moreover
they constitute the watershed for numerous rivers. The Indo-Gangetic alluvium
supports a dense population and has rich potential aquifers. The seismic risk is
also high. The Thar desert is mostly underlain by sediments with potential for
gas and water. Besides, this area is exposed to wind erosion, migration of sand
dunes, spot rains and infrequent flash floods. The upper Cretaceous volcanic
spread or the Deccan traps, as they are usually referred to, are quite extensive and
are spread over an area of 500 000 km2. However, the formation consists of
highly jointed basalts and andesites with the joints persisting to considerable
depth. Vesicles, pores and gas pipes are also met with at depth. Seismic risk is
fair. The Precambrian sediments, that is the Cuddappahs and the Vindhyans, are
composed of sandstones, limestones, quartzites, shales and conglomerates. The
extent of homogeneity in the deposits is unpredictable and also the seismic risk
58 THOMAS et al.
FIG.3. Schematic tectonic map o f southern India.
in the general areas is fair. The coastal plains underlain by Tertiary and Recent
sediments are densely populated, with most of the area under cultivation. The
Precambrian crystalline terrain of the pensinsular shield, on the basis of currently
available data, is predicted to have monolithic formations with the desired degree
of homogeneity and other characteristics. This terrain consists of the Chhotanagpur
area in the State of Bihar, Bundelkhand area covering parts of the States of Uttar
Pradesh and Madhya Pradesh and areas covered by a few adjacent districts of the
States of Andhra Pradesh and Karnataka. Of these, the Chhotanagpur area is
unsuitable due to its high mineral potential and relatively high seismic risk.
As is evident from the above, in spite of the vastness of the area and the
variety of formations, the search for a site for the location of a repository for
high-level and long-lived wastes has to be confined to the Precambrian crystalline
formations in the Bundelkhand area and the southern states.
4.2. Regional geology
A beginning has been made in the investigation of the areas in the Southern
States. The crystalline terrain in parts of Andhra Pradesh and Karnataka States
is under evaluation. The Precambrian shield consisting of these formations has
IAEA-SM-243/158
TABLE III. ARCHEAN SYSTEM OF SOUTHERN INDIA
59
EPARCHEAN INTERVAL
Quartz porphyry and felsite dykes
Closepet or Bellary granites
Chamockites
Peninsular gneisses
Champion gneisses
Dharwar rocks
UNCONFORMITY
Basement — unclassified
remained stable for geological times with little folding or displacement. Fracturing
of the crust in blocks, however, is present due to tension and compresssion.
A schematic tectonic map of the region indicating the major lineations is presented
in Fig.3, which indicates extensive sections of undivided Archean basement
including portions granitized and reworked during Archean/Proterozoic folding.
The Archean system is presented in Table III.
The present investigations for a candidate site for a repository are confined
to the peninsular Gneisses and Closepet granite formations, which are homogeneous
and massive. A schematic geological map of the region is presented in Fig.4. The
peninsular gneisses include gneisses, granites, granodiorites and composite gneisses.
Though the nature of these formations varies from location to location, the
preferred one is the one with uniformly medium or fine-grained gneiss or gneissic
granite. The average dip of the foliation planes of the peninsular gneisses is
moderately steep. However, there is a great variation in both dip and strike and
low dips of 5° to 10° cover considerable sections of the region and at a few
places horizontally lying formations have also been observed. They have well
defined systems of jointing which do not continue beyond a depth of about
100 m.The Closepet granite formation is a band of about 16 km width running
through Closepet, 12° to 15° latitude. They are coarse-grained, porphyrytic
gray-to-pink biotite granites, massive in character.
90 KILOMETRES OnО
LEGEND
DHARWAR SUPER GROUP
PENINSULAR GNEISS
CHARNOCKITE GROUP
CLOSEPET GRANITE
KALAOGI SCRIES
GUDOAPAH SUPER GROUP
KURNOOL SUPER GROUP
DECCAN TRAP WITH MTERTRAPFEANS
ALLUVIUM
FIG. 4. Geological map o f part o f southern India.
TH
OM
AS
et al.
IAEA-SM-243/158 61
Among the sites which are under investigation to assess their suitability for
location of a waste repository a few are indicated in Fig.4. It is to be noted that
all these sites are situated east of Western Ghats. The Ghats lie in the path of the
south-west monsoon and act as a climatic divide. The sites being in the shadow
region are either arid or semi-arid with rainfall around 600 mm.
Relevant information in respect of these sites is summarized.
Site 1 Gray granite — gneiss formations
Site 2 Closepet granites and gray biotite gneisses
Site 3 Gray biotite gneisses
Site 4 , Gray and pink gneisses
4.3. Site identification studies
These locations on the basis of available data do not have any indication of
mineral or other resource potentials. Extensive areas of low relief without any
surface drainage in the vicinity are available. Roads and railways provide good
communications.
Preliminary evaluation of the geotechnical characteristics of the sample rocks
from these sites has been carried out. Though the properties of the rocks vary
from location to location, the variations are limited to a narrow bandwidth. There
is a predominance of felspars in granite gneiss and quartz in granites. Texture in
general is porphyritic. Granites at places show very distinct graphic and myrmekitic
intergrowths. The specific gravity of these rocks varies from 2.63 to 2.71 and
porosity from 0.5% to 1.5%. Young’s modulus is of the order of 5 X 10s kg/cm2.
Compressive strength is of the order of 1600 kg/cm2. Thermal conductivity of
the rocks, on the basis of laboratory evaluation, is of the order of 2 W/m ■ K.
Groundwater conditions in these formations depend on the zone of
weathering and fissures and joints. The surface of water percolates through the
weathered zone which is about 15m thick and is stored in the zone of intermittent
saturation, which is about 3 m thick. The zone of permanent saturation extends
to a depth of about 100 m, where water is stored in fissures and joints. Porosity
of the formation in this zone is about 2%. Beyond this depth the formations
are massive and free from joints. No groundwater has been observed. Further
work on groundwater hydrology is in the planning stage.
It is to be noted that so far no samples of rocks deep below the permanent
saturation zone have been evaluated. In-situ studies on evaluation of thermal
properties of the rocks are also yet to be undertaken.
62 THOMAS et aL
Extensive studies are planned in the locations mentioned earlier so as to
identify promising sites for carrying out pilot disposal studies. The programme
of work includes the following:
(1) Tectonics and seismicity o f the rock formations:
The presently available data indicates that the formations being considered
have been stable for geological times, with hardly any folding or displacement.
However, for purposes of hazard evaluation, detailed studies are planned at the
specific locations under consideration to assure future stability of the formation.
(2) Geological and geophysical studies:
These will include drilling and coring operations, with detailed analysis of the
core.
(3) Groundwater hydrology:
As mentioned earlier, data available to date indicate that there is no ground
water below the zone of permanent saturation, which is at a depth of about 100 m.
Groundwater age, recharging and circulating pattern are some of the parameters
which are under study.
(4) Thermal studies:
Engineering scale mock-up studies to estabUsh the thermal characteristics of
the rocks are planned for the near future. In situ studies will be initiated
thereafter.
(5) Other studies:
Other studies planned include radiation effect on the cores, rock mechanics,
waste-rock interaction, etc.
(6) Hazard evaluation:
A hazard evaluation programme will complement the above studies.
The above-mentioned studies are likely to take about ten years to complete,
leading to the choice of a site for the location of a pilot disposal facility.
4.4. Future programme
IAEA-SM-243/158
BIBLIOGRAPHY
63
BALU, K., RAMACHANDRAN, S., JAIN, B.K., “Management of highly active decladding
zircaloy solid wastes from a fuel reprocessing plant”, Management of Radioactive Wastes
from the Nuclear Fuel Cycle (Proc. Symp. Vienna, 1976) IAEA, Vienna (1976).
GODSE, V.B., et al., “Characterisation of Trombay soils for disposal of radioactive wastes” ,
Disposal of Radioactive Wastes into the Ground (Proc. Symp. Vienna, 1967) IAEA, Vienna
(1967) 301.
GODSE, V.B., et al., “Groundwater Movement Studies in Radioactive Waste Storage Site,
Trombay” — Bhabha Atomic Research Centre Rep. BARC-478 (1970).
SUNDER RAJAN, N.S., KUMRA, M.S., THOMAS, K.T., “Waste immobilisation plant at
Tarapur: ‘A survey of process and design features’, Management of Radioactive Wastes from
Fuel Reprocessing (Proc. Symp. Paris, 1972).
SUNDER RAJAN, N.S., RAVINDRANATH, V., KUMRA, M.S., SINHA RAY, M.K., THOMAS, K.T.,
“Long-Term Planning for Management of Aqueous Wastes from Fuel Reprocessing Plants” ,
Management of Radioactive Wastes from Nuclear Fuel Cycle (Proc. Symp. Vienna, 1976) 1,
IAEA, Vienna(1976) 15.
THOMAS, K.T., et al., “Waste management at Trombay: operational experience”, Management
of Low- and Intermediate-Level Radioactive Wastes (Proc. Symp. Aix-en-Provence, 1970)
IAEA, Vienna (1970).
DISCUSSION
E. PELTONEN: Is the method of storing wastes in trenches and tile holes
at the Tarapur site used only for temporary storage or for final disposal as well?
N.S. SUNDER RAJAN: Currently the storage is considered temporary.
The wastes are retrievable from all the storage facilities although no definite need
for this is indicated at the present stage.
R.H. BECK: In your paper you mention that beyond a depth of about 100 m
the crystalline formations are massive and free from joints. On how many test.
borings is this statement based, and what kind of experiments have you carried out
to verify this important conclusion?
N.S. SUNDER RAJAN: The information provided in the paper is based on
extensive work carried out by the Geological Survey of India and the National
Geophysical Research Institute; it pertains is general to the crystalline complex
of the Precambrian peninsular shield. Test borings are yet to be undertaken at
the sites being considered for location of a waste repository.
H. PIRK: I have a few questions about the dry air-cooled storage facility
for vitrified high-level waste under construction at Tarapur. Is the system based
on convection or forced cooling? How is the waste packaged? And lastly, at
what stage is the project at present?
64 THOMAS et al.
N.S. SUNDER RAJAN: The facility being built at Tarapur to provide
interim storage for solidified high-level waste is essentially an air-cooled vault.
The vault is located underground. The heat removal system is based on natural
convective air-cooling, the convection being induced by a suitably designed stack
and air inlet plenum.
The solidified waste is contained in a canister made of stainless steel 300 mm
in diameter and about a metre in height. The canister is seal-welded. Two such
canisters are placed in a secondary steel containment, which is also seal-welded.
The latter is considered a storage unit.
The engineering design of the facility has been completed. Actual construc
tion should start in a few months’ time.
LAEA-SM-243/95
THE FEDERAL REPUBLIC OF GERMANY’S PROGRAMME FOR THE DISPOSAL OF RADIOACTIVE WASTE
K. KÜHN
Entwicklungsgemeinschaft Tieflagerung,
Clausthal-Zellerfeld,
R.P. RANDL
Bundesministerium für Forschung und Technologie,
Bonn,
H. ROTHEMEYER
Physikalisch-Technische Bundesanstalt,
Brunswick,
Federal Republic of Germany
Abstract
THE FEDERAL REPUBLIC OF GERMANY’S PROGRAMME FOR THE DISPOSAL OF
NUCLEAR WASTE.
The Federal Republic of Germany’s programme goal for the disposal of radioactive waste
is the construction of a repository in the salt dome Gorleben within an integrated “Entsorgungs-
zentrum” . The Federal Agency “Physikalisch-Technische Bundesanstalt” (PTB) is responsible
for this project and has applied for a corresponding licence. The State Government of Lower
Saxony is the licensing authority. In addition to the legally prescribed licensing procedure this
government held a hearing with critics and countercritics of the Federal integrated back-end
of the fuel cycle concept. Following this hearing, Prime Minister Dr. Ernst Albrecht declared
on 16 May 1979 that this concept is technically feasible and in principle acceptable from a
safety point of view. In regard to waste disposal in a salt dome the declaration was also
positive. As a consequence, investigation of the Gorleben salt dome by deep boreholes and
mining exposure will start immediately. The majority of the R&D projects for the disposal
of radioactive wastes in salt formations is carried out by Entwicklungsgesellschaft Tieflagerung
(EGT) and Bundesanstalt für Geowissenschaften und Rohstoffe (BGR). In this regard, the
Asse salt mine plays an important role as a pilot plant and test facility. The carrying out of
R&D studies on radioactive wastes in this mine has recently been complicated by a special new
licensing procedure demanded by the State of Lower Saxony. The German disposal concept
also includes the injection of T-effluents into suitable geological formations and the securing
of separated 85Kr in a surface structure. Corresponding R&D efforts are under way. Further
disposal options being investigated are the cavity technique for low- and intermediate-level
wastes, the use of the abandoned iron-ore mine Konrad as a repository for wastes from nuclear
power plants and decommissioning, and the development of the in situ-solidification process
for low- and intermediate-level wastes in bulk form. The programme of the Federal Republic
of Germany is profiting from increasing international co-operation in waste disposal.
65
6 6 KÜHN et aL
In order to close the back-end of the nuclear fuel cycle, the Government
of the Federal Republic of Germany has developed an integrated nuclear fuel
cycle centre concept, known under its German name as “Nukleares Entsorgungs-
zentrum” (NEZ) [ 1 ]. This centre will consist of the following seven installations:
Storage pools for spent fuel elements
Reprocessing plant
Uranium fuel factory
Production plant for MOX fuel elements
Waste treatment facilities
Repository for radioactive wastes
Facilities for an operational infrastructure.
The Fourth Amendment to the German Atomic Act, 30 August 1976,
declares the Federal Government to be responsible for the disposal of radioactive
wastes. This same amendment appoints the Federal Agency “Physikalisch-
Technische Bundesanstalt” (PTB) in Brunswick to undertake the management of
disposal on behalf of the Federal Government. Therefore, the PTB has to plan,
construct, and operate the repository for radioactive wastes.
Industry is responsible for all other facilities of the “Entsorgungszentrum”.
To fulfil this task, those twelve German utilities operating or planning nuclear
power plants founded a special company, located in Hanover, in 1975. The
“Deutsche Gesellschaft für Wiederaufarbeitung von Kernbrennstoffen (DWK)” is
to plan and operate all other facilities in the nuclear fuel cycle centre, except the
repository [2].
The existence of an appropriate salt dome in which the repository can be
built is one of the basic site requirements for such a nuclear fuel cycle centre. It is
well known that the Federal Republic of Germany has pursued this concept for
the disposal of radioactive wastes for a number of years [3].
Because of its favourable geology, the Federal State of Lower Saxony is
best suited to supply a site for the “Entsorgungszentrum” . On 22 February 1977,
the Prime Minister of Lower Saxony, Dr. Ernst Albrecht, nominated Gorleben
as the preliminary site. Then, on 30 March 1977, DWK started the licensing
process [4]. On 28 July 1977, PTB followed, applying on behalf of the Federal
Government for a licence for installations storing and disposing of the radio
active wastes of the “Entsorgungszentrum” . PTB not only applied for the
emplacement of solid or solidified radioactive wastes, but it also applied for the
injection of tritiated effluents, originating from the reprocessing plant, into
suitable geologic formations in the vicinity of the salt dome and for the securing
of pressurized 85Kr gas flasks in a surface structure. Later in 1978 PTB was
1. INTRODUCTION
IAEA-SM-243/95 67
charged with planning, construction, and operation of an interim storage facility
for high-level glass blocks.
To fulfil its tasks, PTB co-operates with a number of R&D partners, but
mainly the EGT and BGR. EGT (Entwicklungsgemeinschaft Tieflagerung) was
established on 1 January 1978, and consolidates those research and development
activities for the disposal of radioactive wastes performed for many years by the
two national laboratories, GSF (Gesellschaft ftir Strahlen- und Umweltforschung)
and KfK (Karlsruhe Nuclear Research Centre). The second major partner, BGR
(Bundesanstalt fur Geowissenschaften und Rohstoffe), is the Federal Geological
Survey, located in Hanover.
2. DECISIONS IN SPRING 1979
As previously mentioned, the State Government of Lower Saxony chose
Gorleben as the preliminary site for the “Entsorgungszentrum” . At the same
time, this government must also act on behalf of the Federal Government as the
licensing authority for nuclear installations in this State.
After the licence applications of DWK and PTB, the State Government
declared that it would examine both applications very carefully, and that it would
put the safety of people well in the foreground of its considerations.
2.1. Gorleben hearing
In addition to the legally prescribed licensing procedure, the State Govern
ment of Lower Saxony decided independently to hold a hearing in order to
listen to critical opponents of the integrated back-end of the fuel cycle concept.
For this hearing, Lower Saxony selected 62 critics and countercritics of the
concept from home and abroad for a six days’discussion, which took place from
28 March to 3 April 1979, in Hanover in front of a selected and representative
audience. Professor Carl-Friedrich von Weizsâcker, an internationally known
philosopher and scientist for peace, was chosen to be the hearing’s Chairman.
An entire day of this hearing was devoted to discussions of waste disposal
in salt formations. The results of these discussions are reflected in the following
summary of Lower Saxony’s declaration, subsequent to this hearing.
2.2. Declaration of the State Government of Lower Saxony on 16 May 1979
Before the State Parliament on 16 May 1979, the Prime Minister of Lower
Saxony, Dr. Ernst Albrecht, announced the preliminary decision with regard
to the handling of the Gorleben project [5]. The main statement in this
declaration is that the State Government of Lower Saxony considers a “nukleares
6 8 KÜHN et aL
Entsorgungszentrum” to be technically feasible and in principle acceptable from a safety point o f view. This statement was already made on 20 October 1977 by
the Reactor Safety Commission and the Radiation Protection Commission, two
advisory committees to the Federal Government [6]. However, for political reasons the State Government of Lower Saxony recommends that the Federal
Government should not at this time pursue the project of reprocessing.
That part of the declaration which is of particular interest deals with the
disposal of radioactive wastes. Because of the major significance of this statement,
it is quoted here:
“The State Government has satisfied itself that the disposal of radioactive
wastes in a suitable salt dome does not pose any risk to the present generation
and the directly following ones. Also, the risk is small for later generations,
if it is compared to other risks of life.
Because of their plasticity, the salt domes in Northern Germany have
endured more than a hundred million years without their cores being
touched. Several ice ages and catastrophes in the history of earth, such as the
separation of the American-continent from the European continent, failed
to harm them. However, not every salt dome and not every part is equally
suited for the disposal of radioactive wastes. The suitability has to be
examined through careful investigations (boreholes, geophysical investigations,
mining exposure). Scientific and technical methods are available.
With sufficient decay time of the radioactive wastes and with appropriate
spacing it is possible to ensure that the stability of the salt dome is not
affected by the heat production of the high-level wastes.
At most, a risk for future generations would occur if the knowledge of
the disposal of radioactive materials should be lost in the course of centuries
and if later generations should try to mine into the salt dome without knowing
of the repository. But even in this case it has to be pointed out that the
toxicity of repositories with wastes from reprocessing will be drastically
reduced after 500 to 1000 years, so that it can be compared with the toxicity
of natural ore deposits of mercury, lead and uranium.”
As regards disposal, two of four demands which the Prime Minister made at
the end of his declaration are of great importance:
“Push ahead research and development for the safe disposal of radioactive
waste.”
“Drill deep boreholes into the salt dome and, with positive results, explora
tion of the salt dome at Gorleben by mining. In case of negative results
from the boreholes, investigate other sites for disposal.”
2.3. First comment of the Federal Government
IÆA-SM-243/95 69
On the same day, the Federal Government in Bonn made a first comment
on the declaration of the State Government of Lower Saxony. In this comment
important statements were made with regard to waste disposal, so the first
paragraph of this comment is also quoted here:
“The Federal Government appreciates
(a) that the State Government of Lower Saxony considers the
integrated concept of the Federal Government to close the back
end of the nuclear fuel cycle to be technically feasible and in
principle acceptable from a safety point of view,
(b) that the State Government of Lower Saxony accepts, in principle,
the suitability of geological salt formations for the disposal of
radioactive wastes,
(c ) that the State Government of Lower Saxony is willing to start
immediately with the deep boreholes necessary for the investiga
tion of the suitability of the salt dome Gorleben.
The suitability of the salt dome Gorleben is a requirement for the
realization of the concept of the Federal Government to close
the back-end of the fuel cycle. Therefore, the start of the deep
boreholes is an important step towards this objective.”
3. PRESENT CONSEQUENCES
Because the Atomic Act prescribes the “Entsorgung” of nuclear power plants
by reprocessing and waste disposal, and because the Gorleben hearing did not
result in any objections, the Federal Government is adhering to its original concept
of an integrated nuclear fuel cycle centre. However, the present heated discussions
between the Federal Government and all the relevant political and social groups
make more difficult the political acceptance of the integrated concept. Without
being touched by this issue the preparatory projects for waste disposal,are being
continued.
The following five items with regard to radioactive waste disposal reflect
consequences from the events outlined above.
3.1. Site investigations
Investigations of the Gorleben salt dome can be started using the deep
boreholes as soon as drilling permission is granted. Applications were already
70 KÜHN et aL
made in 1977 and 1978. The drillings will be followed by exploratory mining.
In case these investigations should show, contrary to all expectations, that the
salt dome Gorleben is unsuitable, other sites will have to be investigated. These
alternatives are other salt domes in the State of Lower Saxony.
3.2. Time schedule
The time schedule for the repository has not been jeopardized substantially
by the one year delay of the deep-drilling programme. It is still assumed that the
repository will start operation in the early 1990s. The corresponding engineering
planning is under way. A consortium of companies, the so-called “third party
of the PTB”, is being established. Then all the technical and organizational
requirements of Amendment IV will have been fulfilled, enabling planning and
construction of the repository on schedule.
3.3. Increased research and development
As a consequence of the demands by the State Government of Lower
Saxony, the Federal Government will increase current research and development
activities with regard to the disposal of radioactive waste. At present, these
demands are being transformed into appropriate R&D programmes.
3.4. Alternative concept studies
In order to deal with the population’s apprehensions, the Federal Govern
ment agreed to examine other alternative concepts to close the back-end of the
nuclear fuel cycle. This will be done through studies without disturbing the
continuation of the present Federal integrated concept. Because these alternative
concepts could have some impact on waste disposal, additional R&D projects might
have to be performed as a result of these alternative studies.
In order to achieve the same degree of maturity as the present concept the
development of alternative concepts will require at least ten to fifteen years with
a corresponding financial expenditure. It is doubtful, however, if useful alternative
concepts can be generated at all. Nevertheless, the Federal Government has called
upon all critical groups in the country to hand in proposals for alternative concepts.
3.5. INFCE consequences
In pursuing its integrated concept, the Federal Government will make use
of the results of the International Nuclear Fuel Cycle Evaluation (INFCE).
IAE A-SM-243/95 71
The clear goal of the German programme for disposal of radioactive wastes
is to construct and operate a repository in the Gorleben salt dome. This is the
aim of all the site-specific and generic investigations, with the exception of some
peripheral studies in regard to developing disposal operations.
4.1. Gorleben specific projects
The details of the Gorleben salt dome investigation programme, especially
of the deep boreholes and exploration by mining, will be covered in a separate
paper during this Symposium [7].
The architect/engineering conceptual study for the repository was
developed as far as was possible without the availability of site-specific parameters.
This is also covered in that paper.
As already mentioned in Section 3.2, a consortium of three commercial
companies under the supervision of PTB, is being established to deal with the
engineering project of the Gorleben repository. This so-called “third party of the
PTB” will have the name “Deutsche Gesellschaft zum Bau und Betrieb von
Endlagem für Abfalle mbH (DBE)” (German corporation for construction and operation of repositories for wastes, Inc.). The consortium members are
Industrieverwaltungsgesellschaft mbH (IVG), Saarberg Interplan GmbH, and
Salzgitter Maschinen und Anlagen AG (SMAG). The Federal Republic of Germany
has a majority holding in each of the three.
The Reactor Safety Commission (RSK) and Radiation Protection Commission
(SSK), following their declaration of the feasibility in principle of the
“Entsorgungskonzept” from a safety point of view, published a list of recommended
R&D topics and investigations on 18 February 1978 [8]. Since then this list
has been amended to include waste disposal following additional demands from
the licensing authorities of Lower Saxony.
For every R&D topic this list specifies a topic category and a date for
resolution. The category delineates the relative importance which a topic has in
the licensing and advisory procedure, according to the RSK and SSK. The date
for resolution is also a relative value. The use of these relative values is illustrated
by two examples.
One task posed by RSK/SSK reads: “Preliminary specification of a
temperature limit below which changes of camallite can be excluded which could
influence the safety of the repository.”
The relevant category is 2a. This means that both commissions need this
result in order to be able to judge during a preliminary examination if the
respective requirements for licensing are met. The preliminary examination of
the licensing requirements is the date for resolution. According to the time
schedule of PTB, this is 1980.
4. STATE OF THE PROGRAMME
72 KÜHN et a l
Another example is: “Migration of brine inclusions with determination of
migration velocity and of quantities which will be set free. To be followed by
an evaluation of the impacts on safety.” This is a category 1 task, meaning that
RSK/SSK will need these results for their final examination. The date for
resolution is coincident with the detailed planning of the repository, which comes
close to the year 1983.
4.2. R&D projects
Most of the R&D projects, including those on the list of RSK/SSK, have
already been worked on for quite some time. Because of the number and
complexity of these projects they cannot be comprehensively dealt with here.
4.2.1. Asse salt mine
The majority of the R&D projects for the disposal of radioactive wastes in
salt formations is being carried out by EGT and BGR. Both institutions were
mentioned in Section 1. In this regard, the Asse salt mine plays a very important
role as a pilot plant and test facility. Details of the many R&D projects being
performed in the Asse salt mine cannot be delineated here. They are described
in recent publications [4, 9, 10]. As consequences of demands of the licensing
authorities and from evolving political demands (cf. Sections 3.3. and 3.4.),
additional R&D projects will certainly come to the Asse salt mine. At present,
many of these demands have not been transformed into precisely defined
R&D programmes.
The performance of these studies in the Asse salt mine has been further
complicated by a special new licensing procedure demanded by the State of
Lower Saxony. This complication mainly affects the further development of
disposal techniques for low- and intermediate-level wastes. All other R&D projects
continue undisturbed, including the planned test disposal of high-level solidified
waste cylinders, to be retrieved at the end of the test period.
4.2.2. Injection o f T-effluents
In the neighbourhood of the Karlsruhe Nuclear Research Centre a test
injection of T-effluents into an exploited sandstone oil lens is being prepared by
EGT. It is expected that the State agencies of Baden-Württemberg will grant the
licence this year. Extensive experience with the injection of liquid wastes into
geologic formations exists in the Federal Republic of Germany’s oil industry
(injection of many millions of cubic metres of salt water), in the potash industry
(injection of hundreds of thousands of cubic metres of brine), and the chemical
industry (injection of chemical wastes) [11].
IAEA-SM-243/95 73
The discharge of T-effluents into the ocean is considered as an alternative
disposal method. This method is governed by the London Convention of 1975
[12]. In addition, ratification of the Brussels Amendment Treaty is necessary.
This ratification process, is presently under way in the Federal Republic.
Cementing techniques with high degasification and leaching-resistances are
being developed as a second alternative for the disposal of T-effluents.
4.2.3. Securing o f 85Kr
85Kr, which will be separated in the reprocessing plant, is only a short-term
problem of one hundred years. Consequently, the pressurized gas flasks will not
be disposed of in an underground repository, but they will be secured in a
surface structure until the radioactivity has decayed. Correspondingly, the
R&D programmes pursue this objective.
4.2.4. Project “Safety-Studies Entsorgung” (PSE)
The project “Safety-Studies Entsorgung” is dedicated to the complex
compound hazard assessment of all installations within the “Entsorgungszentrum”.
Accordingly, the project is very comprehensive. About 10 institutions are
working together on it. One emphasis of this project is the long-range safety
questions of waste disposal. This project is the subject of a separate paper in
this Symposium [13]. For this reason no details are given here.
4.2.5. Further disposal options
With regard to the disposal of solidified heat-generating high-level wastes
from reprocessing, the goal of the German programme is the emplacement of
these wastes into discrete boreholes, which are to be drilled into the floor of a
repository located in a salt dome.
With regard to the low- and intermediate-level wastes being generated in the
“Entsorgungszentrum” , these will be disposed of in a solidified form in this same
repository. In addition, other technical options are being investigated for both
of these waste categories.
The intermediate-level waste disposal technique developed in the Asse salt
mine uses transportable shielding containers and is very time- and manpower
consuming [9]. Also the throughput is limited. Therefore, EGT has been,working
to develop the so-called “cavity technique” . With this technique the intermediate-
level waste drums are lowered without any shielding underground through a special
shaft and are dropped into a cavity. Construction of the cavity and shaft has been
completed at the Asse salt mine, and installation of handling facilities is nearing
completion. Checking of the total system and cold-test phase will start during
74 KÜHN et aL
this year. Considering the difficult licensing situation of the Asse salt mine, which
was touched upon in Section 4.2.1, it cannot be said at present when the disposal
of radioactive waste can be started.
A second option is being investigated for disposal of wastes from the nuclear
power plants and from decommissioning of nuclear installations. This option
involves the abandoned iron-ore mine Konrad. While Konrad is an iron-ore mine,
this does not mean that Germany is questioning waste disposal in salt. However,
the Konrad disposal concept, like all disposal schemes with geological considera
tions, depends mostly on the unique site-specific geologic and hydrologie
conditions.
The geological setting at Konrad appears very attractive. Another paper in
this Symposium covers this project [14]. The time schedule of the current
feasibility study foresees that a statement in principle will be made in 1980 with
regard to the suitability of the Konrad mine as a repository. In case of a positive
result, the licensing procedure will then be started and the modification of the
present facilities can be initiated. Both these activities may take three to five
years. So radioactive wastes could be disposed of from about 1985.
A third option is being investigated in another R&D project, which studies
the transportation of low- and intermediate-level wastes in a bulk form without
containers. These wastes would be filled into one or several cavities mined in a
salt dome. Finally, these wastes would be solidified in situ in the cavity. This
is an extraordinarily attractive disposal technique and the subject of another
paper at this Symposium [15]. Compared to the two previously mentioned
options, this project of “in situ-solidification” is still in a very early stage of
development so that no date can be given at present for the use of this technique.
4.3. International co-operation
Willingness for international co-operation in the field of radioactive waste
disposal has increased considerably in the last four years. The programme of the
Federal Republic of Germany has been profiting from this fact.
Within the framework of a bilateral agreement for co-operation with the
United States, both partners exchange results without restriction. Also, there is
direct participation of scientists in the mutual programmes. Also, common
R&D projects will be started shortly. In this regard, the Asse salt mine again plays
an important role as a test facility.
Since 1976, a common and harmonized R&D programme for the disposal
of radioactive wastes has existed within the European Communities. Within this
programme, France and the United Kingdom are investigating granite massifs,
Belgium and Italy clay and shale formations, the Netherlands and Germany salt
formations. In this Symposium, another paper discusses the total programme and
its status [16]. The programme is based on a split budget, 50% of which is
IAEA-SM-243/9S 75
financed by the Commission of the European Communities and the other 50%
by the respective Member State. The current programme will be terminated at
the end of 1979, but a second five-year R&D programme is presently being
prepared and will hopefully enable an uninterrupted continuation of the present
programme.
In connection with international co-operation, the OECD/NEA and IAEA
must also be mentioned. Both organizations strive to maintain and to increase
international co-operation. Exchange of knowledge is achieved by their
organizing meetings, symposia, and small workshops which are dedicated to special
topics. In addition, NEA tries to stimulate participation in special national
programmes between its member countries. IAEA has started to elaborate a set
of “Codes and Guides for the Disposal of Radioactive Wastes” [17].
5. SUMMARY AND OUTLOOK
The FRG’s programme goal for the disposal of radioactive wastes is now,
as before, the construction of a repository in the salt dome Gorleben within an
integrated “Entsorgungszentrum”.
The feasibility in principle of this concept from a safety point of view was
confirmed by the declaration of the State Government of Lower Saxony on
16 May 1979.
At present, there are, however, political difficulties regarding the construc
tion of the reprocessing plant at the Gorleben site. In spite of these difficulties,
the politicians also endorse the immediate investigation of the salt dome
Gorleben as to its suitability for radioactive waste disposal. These investigations
into its suitability, by the drilling of deep boreholes, will be started immediately
and will be followed by mining. With positive results from these investigations,
the repository will be constructed in the Gorleben salt dome. In the case of
negative results, another salt dome in Lower Saxony must be explored as a site.
There exists a consensus among politicians, technicians, and scientists with regard
to the necessity and adequate safety of a salt dome repository for radioactive
wastes.
REFERENCES
[ 1] SCHMIDT-KÜSTER, W.-J., The German national fuel cycle policy, Int. Conf. Nuclear
Fuel Cycle, London, 1978, Atomic Industrial Forum; British NucL Forum (1978).
[2] SALANDER, C., The concept of the German electric power industry for the disposal
of spent fuel from nuclear power plants, Kemtechnik 20 (1978) 229—237.
[3] KÜHN, K., Geoscientific investigations in the Asse II salt mine, Disposal of Radioactive
Wastes into the Ground (Proc. Symp. Vienna, 1967), IAEA, Vienna (1967) 509—518.
76 KÜHN et aL
[4] KUHN, K., RÔTHEMEYER, H., SALANDER, C., West Germany gears up for licensing
process, Nucl. Eng. Intern. 23 (1978) 48—53.
[5] Regierungserklàrung von Ministerprasident Dr. Ernst Albrecht, 16. Mai 1979, Pressestelle
der Niedersàchsischen Landesregierung, Hanover (1979).
[6] Grundsâtzliche sicherheitstechnische Realisierbarkeit des Entsorgungszentrums. Beurteilung
und Empfehlungen der Reaktorsicherheitskommission (RSK) und der Strahlenschutz-
kommission (SSK), Bundestags-Drucksache 8/1281 dated 30.11.1977, 12—35.
[7] RÔTHEMEYER, H., “Site investigations and conceptual design for the mined repository
in the nuclear “Entsorgungszentrum” of the Federal Republic of Germany” , these
Proceedings, SM-243/48.
[8] Sicherheitstechnische Fragestellungen zum Entsorgungszentrum. Stand der Beratungen
der RSK und der SSK. Empfohlene F&E Arbeiten und Untersuchungen, Geschâfts-
stelle der Reaktorsicherheitskommission, Cologne, 18.2.1978.
[9] KÜHN, K., et aL, “Recent results and developments on the disposal of radioactive wastes
in the Asse Salt Mine” , Management of Radioactive Wastes from the Nuclear Fuel Cycle,
(Proc. Symp. Vienna 1976) 2, IAEA, Vienna (1976).
[10] KÜHN, K., HAMSTRA, J., “Geologic isolation of radioactive wastes in the Federal Republic
of Germany and the respective program of the Netherlands”, Management of Wastes from
the LWR Fuel Cycle (Proc. Int. Symp. Denver, 1976, CONF-76-0701 (1976)) 580 —600.
[11] AUST, H., KREYSING, K., Geologische und geotechnische Grundlagen zur Tiefversenkung
von fliissigen Abfailen und Abwassem, BGR-Forschungsbericht 10301 001, Hanover,
(1978).
[12] UN Convention on the prevention of marine pollution by dumping of wastes and other
matter, “The definition required by annex I, paragraph 6 to the Convention and the
recommendations required by annex II, section D”, IAEA, INFCIRC/205/Add.l,
10 January 1975.
[13] LEVI, H.W., Project “Safety-Studies Entsorgung” in the Federal Republic of Germany” ,
these Proceedings, SM-243/17.
[14] BREWITZ, W., LOSCHHORN, H., “Geoscientific investigations in the abandoned iron-ore
mine KONRAD for safe disposal of certain radioactive waste categories”, these
Proceedings SM-243/14.
[15] KÔSTER, R., KROEBEL, R., KRÀMÉR, R., “Disposal and fixation of medium- and
low-level liquid wastes in salt caverns” , these Proceedings SM-243/16.
[16] VENET, P., DELLA LOGGIA, E., FALKE, W., “The European Communities’ R&D
programme on the disposal of radioactive wastes into geological formations: progress
and results” , these Proceedings SM-243/128.
[17] Site Selection Factors for Repositories of Solid High-Level and Alpha-Bearing Wastes
in Geological Formations, Technical Reports Series No. 177, IAEA, Vienna (1977).
DISCUSSION
H.W. LEVI: I should like to comment on the political reasons which the
Prime Minister of Lower Saxony referred to when he proposed the postponement
of reprocessing as part of the integrated concept of spent fuel and waste manage
ment (Entsorgung). These are reasons connected with the lack of public
acceptance in the Federal Republic of Germany rather than what are usually
considered to be political factors.
IAEA-SM-243/95 7 7
K. KUHN : These political reasons are a combination of public acceptance
considerations and real political factors. For, if there are differences between
different political parties with respect to acceptance, how can the man in the
street be expected to accept a complicated nuclear complex Uke the spent fuel
and waste management centre (Entsorgungszentrum)? After all, he feels that
he is represented by the politicians whom he elected to Parliament.
G. ROCHLIN: If I understand you correctly, the Federal Republic of
Germany — unlike the USA, the Soviet Union, Finland and India, all of which are
seeking to expand their programmes of exploring many options — does not plan
to study the possibility of spent fuel disposal at all, even of small amounts.
I gather that it does not intend to investigate alternative geologies or media at all
and that it will not explore several domes at once but turn its attention to a second
or third choice only if the preceding one proves unsuitable.
I wonder if you would comment on the contrast between the increasing
diversity of exploration of alternatives in other countries and the continued focus
of the FRG’s programme on a single site which was not selected according to any
of the basic screening criteria described at this meeting and does not satisfy some
of them.
K. KÜHN: I should like to reiterate that the Federal Government adheres
to an integrated concept which includes reprocessing and for that reason we are
not considering disposal of spent fuel elements.
The clear goal of our programme is the construction and operation of a
repository in the Gorleben salt dome. We are not ourselves investigating
alternative geologic formations for waste disposal. However, in accordance with
our agreement with the Commission of the European Communities, we have
unlimited access to all information obtained in other Member States of the
Community on other formations. Furthermore, under a bilateral co-operation
agreement between the USA and the FRG we also have access to American findings.
The suitability of the Gorleben salt dome as the site for a repository will be
carefully investigated. If these investigations yield negative results, another salt
dome in Lower Saxony will be examined. Now, as before, we are concentrating
on salt.
H. KRAUSE: I should like to comment briefly on Mr. Rochlin’s remarks.
I don’t think it would be advisable for each country to investigate all possible
approaches to final disposal. Such a policy would involve a strain on resources
and there is a risk that in the end no real progress would be made. In my opinion,
it is better to concentrate on one or two concepts, trying to obtain detailed data
in the shortest possible time and using this information for the design of suitable
facilities. The sum of all these activities will certainly be more valuable than a
wide scatter of activities in each country.
G.E. COURTOIS: In an INFCE report I read that the investment on a
radioactive waste repository to support a 50 GW(e)/a economy for 30 years
78 KÜHN et al.
would be DM 5000 million. Could you confirm this figure? Is it intended in the
FRG that all low-, medium- and high-level wastes should be disposed of in
geological formations?
K. KÜHN: INFCE report WG 7 is based on the artificial case of a repository
(either salt or granite) used for the disposal of all types of radioactive waste from
different nuclear fuel cycles, with the exception of mining and milling wastes.
So you have to take the figure in its proper context.
The answer to the second question is yes, although the option of dumping
tritium waste or effluents and separated 85 Kr into the sea is under consideration.
C.A. HEATH: I would like to amplify the reply given by Mr. Kuhn concerning
the use of cost figures from the INFCE studies. The values calculated should be
used only to make a relative comparison between the various fuel cycles studied,
since several very conservative assumptions were made for purposes of comparison.
For the sake of agreement among many countries on the basis for comparison,
for example, all wastes arising from the fuel cycle with the exception of mining
and milling wastes were assumed to be placed in a deep geologic repository.
This assumption meant, therefore, that low-level wastes, often assigned to
shallow land burial and even the depleted uranium from the enrichment process
were included for geologic disposal. Because it is most unlikely that all these
wastes would go to a geologic repository, the estimated cost used in the INFCE
study for geologic repositories to support a 50 GW(e)/a economy should not be
used as an absolute figure.
IAEA-SM-243/167
DEVELOPMENT OF DEEP UNDERGROUND DISPOSAL
FOR CANADIAN NUCLEAR FUEL WASTES
S.R. HATCHER, S.A. MAYMAN, M. TOMLINSON
Atomic Energy of Canada Limited,
Whiteshell Nuclear Research Establishment,
Pinawa, Manitoba, Canada
Abstract
DEVELOPMENT OF DEEP UNDERGROUND DISPOSAL FOR CANADIAN NUCLEAR
FUEL WASTES.
The programme of research and development for safe disposal of fuel wastes from nuclear
power production in Canada is described and the current status outlines. Immobilization
technology is being developed so that either used fuel or separated fuel wastes can be isolated
deep in plutonic igneous rock formations. The development programme, involving several
government departments, universities and industry, is expected to take some twenty years. The
first phase is aimed at verifying the basic concept of deep underground disposal in hard rock of
the Canadian Shield. It is expected that, by late 1981, selection of a technically suitable site
for a demonstration disposal facility will be possible. Construction of the facility will follow and
operation is expected in the late 1980s. Throughout the programme considerable emphasis is being
given to providing full and open information to the public.
1. INTRODUCTIONThe overall objective [1] of the Canadian radioactive
waste management program [2-4] is to ensure that there will be no significant effects on man or his environment at any time. This objective has two aspects:
Safety - to manage the radioactive by-products and wastes so that the potential hazards are negligible.Responsibility - to manage radioactive by-products and wastes in such a way that the trouble and concern to future generations in keeping them safely will be minimized or eliminated.
Wastes from the nuclear fuel cycle vary considerably in terms of quantity, level of radioactivity and half-life of the principal radionuclides contained. Safe management of particular types of waste, for example, mine/mill tailings, reactor operations wastes or used-fuel wastes, can therefore be achieved in a variety of ways. In the case of used-fuel wastes, Canada, like
79
80 HATCHER et al.
many other nations, is concentrating on disposal deep underground in stable geologic formations.
The purpose of the research, development and demonstration program outlined here is to verify that disposal of used- fuel wastes in deep, stable formations will achieve the overall objective and to establish the capability for safely disposing of these wastes. Some consideration is being given to the disposal of immobilized reactor wastes and the wastes which will be produced in reactor decommissioning in the same facility. However, since the largest quantities of radioactivity and heat generation are associated with used-fuel wastes, it is these which will control both the design and performance of any underground facility.
Atomic Energy of Canada Limited (AECL), a federal Crown Corporation, has the prime responsibility for this research and development program. An agreement between the governments of Canada and Ontario provides for full consultation and cooperation in the management of nuclear fuel wastes. As part of this agreement, Ontario Hydro, the provincial utility, is undertaking studies of necessary pre-disposal technologies - the interim storage and transportation of used fuel.
The program covers a wide range of scientific and engineering disciplines. To provide the necessary expertise, several departments and agencies of the governments of Canada and Ontario are participating in the program. Major contributions are also being made by the university community and by private industry.
2. APPROACH
The development principle throughout the Canadian nuclear program has been to identify the most attractive approach for Canada and concentrate the available resources on that approach. This principle is followed in the nuclear fuel waste management program,which is concentrating on disposal in the stable, hard-rock formations of the Canadian Shield which underlies much of the country [2]. Work on other formations, such as bedded salt, has been limited to the identification of potentially suitable deposits. However, by participating in the world-wide exchange of information on waste management programs, in international studies such as the Seabed Working Group, and by keeping abreast of work in other countries, Canada maintains the option of pursuing disposal in other geologic formations should hard- rock disposal prove unattractive.
IAEA-SM-243/167 81
The approach being taken is also influenced by the features of the C A N D U 1 reactor system and by the national nuclear power program. Operating on a natural uranium, once-through fuel cycle, excellent uranium utilization is achieved and no credit is taken for the fissile plutonium contained in the used fuel.There is, therefore, no immediate need, on economic or other grounds, to recover and recycle the fissile material. In Canada, no decision has yet been taken on the future recovery and use of plutonium. The government has stated that its position on fuel recycle will take into account the results of the International Nuclear Fuel Cycle Evaluation now in progress. Current used-fuel storage methods are adequate for many years, and additional storage capacity can readily be provided at the reactor sites or at a central location. Thus, commercial disposal is not an immediate requirement. Nevertheless, recognizing the eventual need for disposal, technologies are being developed for the immobilization of both used fuel and reprocessing wastes so that options are kept open for the disposal of either form. It is intended to take each at least to the pilot plant stage.
The Canadian development program [2], expected to take some twenty years, has three phases - concept verification, site selection and the construction and operation of a demonstration disposal facility.
Phase 1 - Concept Verification
This research-oriented phase is aimed at verifying the basic concept that effective isolation of the radioactive material from man and the environment can be achieved by disposal of immobilized fuel or reprocessing wastes some 500 to 1000 metres deep in stable, hard-rock formations.
The hard-rock formations known as plutons2, cover a wide spectrum of rock types and fracture patterns. Laboratory and field research studies are being made to determine their properties and hydrogeologic characteristics. These data, together with information on the behaviour of immobilized wastes and the sorption properties of buffer and backfill materials, will be used to assess the effectiveness of the various natural and engineered barriers to the movement of radioactive materials. Development of the methodologies to make such an assessment is a vital part of the program, and is described in a companion paper[5].
1 CANada Deuterium Uranium.
2 Pluton: typically a large body of intrusive igneous rock.
82 HATCHER et al.
Although the research work on all aspects will continue for many years, it is expected that by late 1981 it will be possible to identify which of the various types of pluton will be technically suitable for a waste disposal facility. Furthermore, it is expected that at that time sufficient data will be available to make an assessment of the environmental impact of such a facility, and to demonstrate with reasonable confidence that the basic concept is sound.
Phase 2 - Site Selection
The results obtained in the first phase will allow the selection of a large number of technically suitable sites among the 1500 plutons which have been identified in the Ontario portion of the Canadian Shield. The site selection phase will explore the social and political considerations associated with these technically suitable sites. Specific information on the proposed facility, including details on economic factors, community impact, health and safety considerations and environmental effects, will be discussed with communities near these sites.The elected representatives of the communities will participate in the decision on whether the facility will be built in their area.
Phase 3 - Construction and Operation of the Demonstration Facility
Once a site has been acquired and its technical suitability confirmed, a demonstration facility will be constructed. The facility is not expected to be operational until the late 1980's. The facility will resemble a typical hard-rock mine.
Initial testing in the facility will not involve radioactive materials. Testing of equipment and handling methods will be done along with the measurement of physical characteristics such as rock stress and heat transfer capability. Once the thermal and mechanical design of the facility has been confirmed, immobilized wastes will be emplaced and an extensive monitoring program initiated. Should the waste form or location prove unsuitable, the wastes would be retrievable during this phase of the demonstration.
Provided that the performance is satisfactory, the facility could be expanded to a commercial operation. Alternatively, a commercial facility could be built at another location. However, a decision on commercial disposal is not likely before the end of this century.
IAEA-SM-243/167 83
Public understanding and acceptance is important to the success of the program. Considerable emphasis is being given to providing full and open public information. In particular, the concept verification phase allows an opportunity to increase public awareness of the program. This will lay the foundation for public involvement during the site selection and subsequent phases. The public, through their elected representatives at all levels of government, will be involved in any decisions relating to their community. The regulatory and licensing processes which must be followed also provide for public participation through submissions and hearings.
A Technical Advisory Committee to AECL on the nuclear fuel waste management program has been appointed. This Committee, consisting of distinguished Canadian scientists, is asked to review all aspects of the program in detail and to advise on them. Their comments and recommendations will be freely available to the public.
3. SUMMARY OF THE R & D PROGRAM
Although interim storage and transportation of used fuel are important aspects of the overall used-fuel waste management program, this paper covers only the work directly involved in development and demontration of underground disposal. A summary of the components of the program and the schedule for these activities is given in Figure 1.
3.1 FUEL IMMOBILIZATION
The objective of the fuel immobilization program is to develop additional containment for safe disposal of unreprocessed fuel. The fuel matrix already provides a stable, solid matrix of low solubility in water. Most of the radioactive material is effectively trapped in this matrix. The Zircaloy cladding adds another barrier to the escape of fission products.
The program is concentrating on two approaches. One is the development of a system which will give a high probability of containment for at least 300 years, thus ensuring isolation of nearly all fission products for their hazardous lives; the other is to investigate concepts which might offer the prospects of a substantially longer period of isolation.
In the first approach, a shell of a corrosion-resistant metal is being considered as the major barrier to the release of radioactive material. Several metals or alloys are potential
O V E R A LL PROGRAM
FUEL IM M O B ILIZA TIO N
WASTE IM M O BILIZATIO N
GEOTECHNICAL RESEARCH
V A U LT DESIGN & CONSTRUCTION
SAFETY & E N VIR O N M EN TAL ASSESSMENT
SUPPORTING RESEARCH
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Ш ШXCOMNCONCEPT
VERIFIC ATIO NSITE
vSE LECTIONDEMO. V A U LT CONSTRUCTION
TEST, L O A D & EVALU ATE 'M M ERCIALO ADING
ÛEVELOP X BU ILD PILOT \ PRODUCE TEST LO AD§IMPI„g C Q N I^ I p la n \J___________X _________________
DEVELOP AD VANCED CONTAINER___________ \Ш4 BU ILD D E M O X PRODUCE & TEST
_________ /\ PRODUCE/ACQUIRE TEST FORMS X
BU ILD
A D V A N C E D CONTAI N E F ^ p u f f i p
DEVELOP WASTE FORMS EXPERIM ENTAL PRODUCTION
TEST & E V A LU A TE UN DERG RO UND
BU ILDLARGERPLANT
D R IL L & INVEST- X STUDY SELECTED IGATE FORMATIONS SITE
STU D Y H Y D R O G E O LO G Y .________________________________ ____________________^ M O N IT O R M EC H AN IC AL & T H E R M A L EFFECTS IN DEMO V A U LT
CONCEPT DEVELOPMENT
V A U LT _ DESIGN
VSINK \E X C A V A T E \S H A F T F V T E S T VAU
FU RTH ER DEVELOPMENT BA C K FILLIN G & SEALING TESTS
FURTHER EXCAVATION
- ЩLICENSE, A P P R O V É X
OM M ER CIAL LO AD IN G
ШDEVELOP \U C E N S jN G & _A P P R O V A L S SITE_& JTE^TS’ C O N T I N U E TO EVALUATE, REFINEM ETHODOLOGY REFINE M ODËlJ ~ & TEST MODELS
LA BO R ATO R Y SIM ULATIO NS PROVIDE BASIC D A TA A N D TEST MODELS
" V LICEV o
DEVELOP BASIC UNDERSTANDING & D A TA IN GEOLOGY, HYDROGEOLOGY, CHEMISTRY, M A T E R IA L SCIENCE, EN VIR O N M EN TAL SCIENCE ETC
------------------------ --F ^ l ---------------------------
FIG.l. Schedule: Development program for Canadian nuclear fuel waste disposal.
HAT
CH
ER
et al.
IAE A-SM-24 3/167 85
candidates, depending on the chemistry of the water passing through the disposal facility. Additional materials in the container, which retard radionuclide release, can also be included. The program calls for a pilot plant to be available for making representative containers of immobilized fuel by the mid-1980's.
The second approach is being addressed on a longer time-scale. A containment providing assurance of long-term isolation will be more complex, involving a series of barriers. Encapsulation in metals or ceramics with metal matrixing and the incorporation of retardants to the movement of specific radionuclides is under study. •
3.2 FUEL WASTE IMMOBILIZATION
The objective is to develop the capability to incorporate the fission products and associated wastes from the reprocessing of CANDU used fuel into a low-solubility matrix. The program is directed primarily toward incorporation of wastes in conventional borosilicate glasses, but attention is also being given to other matrices, including crystalline solids. Consideration is being given to incorporating virtually all the active liquid waste streams into one glass with a low fission-product loading. This implies the production and handling of a larger number of glass blocks; however, this disadvantage may be compensated by lower heat loadings and temperature gradients and a reduction in the variety of waste products. Work on the more conventional high fission-product loadings is also being done. Efforts are also directed toward the development of appropriate matrices for the immobilization of volatile reprocessing wastes such as 129I, 85Kr, 14C and 3H.
To provide active waste streams for test and development purposes, use will be made of laboratory-scale facilities for experimental processing of thoria fuel, and of facilities being developed to immobilize high-level liquid residues from medical isotope production operations. As with fuel immobilization, a number of immobilized fuel waste capsules will be required by the late 1980's. Testing of some active waste forms from sources outside Canada could be of interest if suitable arrangements can be made.
3.3 GEOTECHNICAL RESEARCH
The first objective of the geotechnical research is to provide an information base from which technically suitable locations for a demonstration disposal facility can be selected. The geologic formations have been classified into four general rock types: granite, syenite, anorthosite and gabbro. Since
8 6 HATCHER et al.
groundwater chemistry and flow will be influenced by rock type and degree of fracturing, the research will examine samples of all four rock types with a wide range of fracture densities.
Information will be obtained on the rock composition and properties, rock homogeneity, degree and nature of fracturing within the rock mass, and particularly on the characteristics of the groundwater flow and its flow system. Most of this information can only be obtained from detailed field investigations, supplemented by laboratory examinations of core samples.
The geologic investigations are being done by various branches of Energy, Mines and Resources Canada and the hydrogeological work in large measure by Fisheries and Environment Canada. The geochemical studies are done at several universities and in AECL's laboratories. Some of this work is described more fully in another paper [6] at this conference.
3.4 PRELIMINARY DISPOSAL VAULT DESIGN
Preliminary design work is being done to investigate a range of underground configurations. It has identified layout and construction features of the facility, methods of emplacing the immobilized waste and methods of backfilling and sealing such a facility. Detailed thermo-mechanical analyses of reference configurations are important features of the design. The work is being done by mining engineering consultants and is summarized in a companion paper [6] at this conference.
3.5 SAFETY AND ENVIRONMENTAL ASSESSMENT
The key to acceptance of the disposal concept by regulatory and licensing authorities and by the public is proof of the safety of the system. Methodologies and associated analytical tools are being developed to perform the necessary safety and environmental assessments. The possible migration of radionuclides from the emplaced immobilized waste through the geologic formation to the biosphere, and the possible radiation dose to future populations, will be analyzed to form the basis of these assessments.
Preliminary models are almost complete for the various zones and modes of migration of radionuclides; advanced methodologies are being developed. Some aspects of this work are covered in a separate paper [5] at this conference.
It is also necessary to demonstrate that the component models are valid and to obtain the basic physical and chemical data required for their use. Experimental programs have been
IAEA-SM-243/167 87
u n d e r ta k e n to a c c o m p lis h t h i s . They a re s t u d y in g v a r io u s a s p e c ts
o f th e w as te - ro ck - g ro u nd w a te r i n t e r a c t i o n and a ls o th e e f f e c t i v e
n e ss o f b u f f e r and b a c k f i l l m a t e r i a l s . The w ork in c lu d e s meas
urem en ts o f th e movement o f w a te rb o rn e r a d io n u c l id e s and th e
t r a n s fo r m a t io n s o f m a t e r ia l s and s u r fa c e s i n l a b o r a t o r y system s
c o n t a in in g w a te r , r o c k , w as te and o th e r m a t e r i a l s . Work on a
l a r g e r s c a le i s p la n n e d and f i e l d t e s t s w i l l f o l lo w .
3 .6 SUPPORTING RESEARCH
T h is work h as th e o b je c t i v e o f im p ro v in g th e u n d e r
s t a n d in g o f th e s c ie n c e u n d e r ly in g th e deve lopm en t o f deep u n d e r
g round d is p o s a l t e c h n o lo g ie s . The a re a s o f i n t e r e s t co v e r g e o l
o g y , c h e m is t r y , m a t e r i a l s s c ie n c e , e n v ir o n m e n ta l s c ie n c e and
o t h e r s . Some o f th e q u e s t io n s b e in g ad d re sse d a r e : th e mecha
n ism s o f s o r p t io n on r o c k s ; th e rm a l and h y d ro th e rm a l t r a n s fo rm a
t i o n o f g la s s e s and r o c k s ; f r a c t u r e p r o p a g a t io n i n r o c k ; th e up
ta k e o f r a d io n u c l id e s by f l o r a and fa u n a o f th e C a n a d ia n S h ie ld .
4 . STATUS OF TECHNICAL PROGRAM
I t i s n o t p o s s ib le to p r o v id e d e t a i l s o f th e t e c h n ic a l
p ro g re s s i n a n a t i o n a l p rog ram o f th e b r e a d th d e s c r ib e d i n t h i s
p a p e r . Many v i t a l a c t i v i t i e s a r e i n p ro g re s s w h ich w i l l n o t be
m e n tio n e d i n t h i s p ap e r o r i n o th e r c o n t r ib u t io n s to t h i s con
fe r e n c e . They a re d e a l t w i t h i n o th e r r e p o r t s and p u b l i c a t i o n s
[ 2 ,7 ] .
The i n t e r a c t i o n o f th e w as te fo rm w ith th e g e o lo g ic
e n v iro n m e n t i n w h ich i t i s p la c e d i s th e fo c u s o f th e p rogram in
b o th f i e l d and la b o r a t o r y r e s e a rc h and i n th e a sse ssm en t s t u d ie s .
E f f e c t s o f g ro undw a te r on w as te fo rm s a r e b e in g exam ined b o th i n
th e f i e l d and i n th e la b o r a t o r y . Two o f th e n e p h e lin e s y e n i t e
g la s s b lo c k s c o n t a in in g r e p r o c e s s in g w a s te , w h ich have been b u r
ie d a t C h a lk R iv e r f o r some 20 y e a rs [8 ] , h ave been re c o v e re d and
a r e unde r d e t a i l e d s c r u t in y a t th e W h it e s h e l l N u c le a r R ese a rch
E s ta b l is h m e n t (WNRE). No s i g n i f i c a n t d e t e r io r a t i o n h a s been de
t e c t e d . S in c e th e lo n g - te rm le a c h r a t e s m easured i n th e f i e l d
a r e v e ry much low e r th a n th o s e p r e v io u s ly m easured i n l a b o r a t o r y
e x p e r im e n ts , sam p les o f th e re c o v e re d g la s s b lo c k s a r e b e in g
te s te d to in v e s t ig a t e th e re a s o n f o r th e s e d i f f e r e n c e s [9 ] . Leach
r a t e m easurem ents o f UO2 i n s t a t i c , ox ygena te d w a te r a t room tem
p e r a tu r e have been i n p ro g re s s f o r a lm o s t t h r e e y e a r s . The r a t e
has d e c re a se d w i t h t im e and i s now le s s th a n 3 x 10~? g U0£ pe r
sq u a re c e n t im e t r e p e r d ay .
88 HATCHER et al.
100 1000 1 0 000
T IM E A F T E R E M P L A C E M E N T (YEARS )
FIG.2. Mid-point temperature rise in vault.
Many of the interactions between waste forms, water, geologic medium and leached radionuclides can be affected or controlled by physical or chemical means. Effort is therefore being devoted to developing backfill and buffer materials which can retard the ingress of water, adjust the chemistry of the water which does pass through to reduce the rate at which it attacks the waste form, and act as a scavenger for any radionuclides which are leached from the wastes. The reference material at this time is a sand-bentonite mixture. However, others are being considered, either alone or in combination, which will not limit temperatures to the extent that bentonite does and which will be more effective in removing ^ T c 1 2 9 ^ Lead titanate shows some promise in the latter regard.
The preliminary design of the disposal facility itself is covered in some detail in a companion paper [6]. Heat produced by the wastes seems to be the only factor which distinguishes the design analysis for this facility from that required for a conventional mine, although handling and emplacement techniques suitable for handling highly radioactive sources must be developed. A significant difference has been noted in the design
IAEA-SM-243/167 89
constraints for immobilized fuel compared to those for reprocessing wastes. As shown in Figure 2, elevated temperatures in the vicinity of the vault persist much longer for fuel disposal than for waste disposal. To ensure that effects in the far-field are similar, peak temperatures in the fuel vault may have to be lower than those permitted with the waste facility. In present designs this results in almost a factor of two increase in the size of the vault facilities.
A central activity in the concept verification phase of the program is the geotechnical investigation of plutonic igneous bodies in the Canadian Shield. The work is described in another paper [6]. It has generally been concentrated within the Province of Ontario, though some drilling has been done elsewhere. Almost 1500 formations have been identified, with a rough distribution of granite (75%), syenite (8%), anorthosite (2%) and gabbro (15%). Research areas are being selected with due regard to rock type, size, exposure at the surface, degree of faulting and fracturing, and ease of access.
The AECL research laboratories at both Chalk River and Whiteshell are located on crystalline hard-rock formations and airborne, surface and borehole testing has now been in progress at these sites for over two years. The Chalk River formation, adjacent to the Ottawa Valley fault, is highly fractured and deformed. Nine boreholes have been drilled there, the deepest to 700 metres, and a number of geophysical tools and techniques have been tested and calibrated. Whiteshell is on the large, comparatively unfractured, Lac du Bonnet batholith. Four boreholes have been completed there, the deepest to 900 metres, and these are being used primarily to develop hydrogeologic methods and equipment. Rather wide zones, up to 200 metres, essentially free of fractures have been encountered in the Lac du Bonnet batholith, but some well-fractured zones exist as well. Several of the boreholes were logged in late 1978 by a United States Geological Survey crew, and similar cooperative efforts are expected in the future.
Data from all test programs, and the models developed to correlate those data, are employed in the preparation of assessments of the environmental impact of the entire waste management program. The first rough pathway analysis for a disposal facility in hard-rock has been completed [5]. The general conclusion at this time is that the multiple barriers can provide sufficient protection for man and the environment. More refined assessments, based on a better understanding of the processes and systems providing this protection, are expected to demonstrate the conservatism of present analyses.
90 HATCHER et aL
REFERENCES
[1] DYNE, P.J., Waste Management in Canadian Nuclear Programs, Atomic Energy of Canada Limited Report AECL-5249 (1975).
[2] BOULTON, J. (ed.), Management of Radioactive Fuel Wastes: TKe Canadian Disposal Program, Atomic Energy of Canada Limited Report AECL-6314 (1978).
[3] MAYMAN, S.A., BARNES, R.W., GALE, J.E., SANFORD, B.V.‘,‘The Canadian program for storage and disposal of spent fuel and high-level wastes” , Management of Radioactive Wastes from the Nuclear Fuel Cycle (Proc. Symp. Vienna,1976) Vol.l, IAEA, Vienna (1976) 49.
[4] TOMLINSON, M., et al., “Management of radioactive wastes from nuclear fuels and power plants in Canada” , Nuclear Power and its Fuel Cycle (Proc. Int. Conf. Salzburg, 1977), IAEA, Vienna (1977) 167.
[5] LYON, R.B., ROSINGER, E .L J., “ Safety assessment for deep underground disposal vault — pathway analysis” , these Proceedings, SM-243/169.
[6] SCOTT, J., CHARLWOOD, R.G., “ Canadian geoscience research and design concepts for disposal of high-level waste in igneous rocks” , these Proceedings, SM-243/168.
[7] HAWLEY, N.J., Radioactive Waste Management in Canada: A Bibliography of Published Literature, Atomic Energy of Canada Limited Report AECL-6186 (1978).
[8] MERRITT, W.F., “The leaching of radioactivity from highly radioactive glass blocks buried below the water table: Fifteen years of results” , Management of Radioactive Wastes from the Nuclear Fuel Cycle (Proc. Symp. Vienna, 1976) Vol.2, IAEA, Vienna (1976) 27.
[9] STRATHDEE, G.C., McINTYRE, N.S., TAYLOR, P., “Development of aluminosilicate and borosilicate glasses as matrices for CANDU high-level waste” , Ceramics in Nuclear Waste Management (Proc. Int. Symp. Cincinnati, 1979), Paper 51-SI-79 (to be published).
DISCUSSION
K. KÜHN : In Section 4 of your paper you refer very briefly to nepheline
syenite glass blocks which had been in tbe ground at Chalk River Nuclear Labora
tories for twenty years and were then recovered and transported to the WNRE.
What types of examination were made on these specimens and what were the
results?
S.R. HATCHER: The studies included measurements of density and radia
tion fields, 7-spectrometry, X-ray photoelectron spectroscopy, scanning electron
microscopy and energy dispersive X-ray spectrometry. One sample indicated
calcium depletion to a depth of 0.3 mm from the surface, but it was considered
probable that this resulted from the fabrication process. X-ray photoelectron
spectroscopy revealed depletion of some elements from the surface to a depth of
two nanometres in all surface specimens of both glass blocks.
G. STOTT: You rightly mention the need for public participation in disposal
strategy. Was there any public acceptance problem in connection with the drilling
work being carried out at Whiteshell?
IAEA-SM-243/167 91
S.R. HATCHER: The drilling at Whiteshell is an extension of the ongoing
nuclear power research at the establishment and posed no problems of public
acceptance. Public information and community relations programmes are under
way in several regions of Ontario and have led to the approval of a new research
area near Atikokan in the north-west of the province. The approval followed
several months of information meetings between AECL scientists, politicians and
the public. All three levels of government, local, provincial and federal, have
given their approval for research drilling.
Nina KRYLOVA: Is it intended to vitrify wastes of all levels of activity?
Will wastes with different levels of activity be buried together or separately?
S.R. HATCHER: Vitrification is planned for wastes which would be produced
if the fuel were reprocessed. Other low- and medium-level wastes from reactor
operations may be concentrated and bituminized. These wastes may be disposed
of in a separate facility or in the fuel wastes vault.
H. KRAUSE: What will be the Canadian policy with regard to alpha-bearing
wastes in the case of reprocessing? Is it intended to bury such waste at shallow
depth or will it be disposed of, together with the high-level wastes, in a deep
geologic formation?
S.R. HATCHER: It is expected that such alpha-bearing wastes would be
disposed of in the fuel wastes vault in a deep geologic formation.
IAEA-SM-243/160
CONCEPT AND REALIZATION PROGRAMME FOR
FINAL STORAGE OF NUCLEAR WASTE
IN SWITZERLAND
H. ISSLER, R.H. BECK, H. ZÜND
Nationale Genossenschaft für die Lagerung
radioaktiver Abfalle NAGRA,
Baden, Switzerland
Abstract
CONCEPT AND REALIZATION PROGRAMME FOR FINAL STORAGE OF NUCLEAR WASTE IN SWITZERLAND.
The Swiss concept for nuclear waste disposal is briefly discussed with emphasis on the classification of waste and the selection of appropriate host rocks for the various types of repositories. The legal link created by the passing of the revised Atomic Law between the licences for new nuclear power plants and waste repositories is explained. The major elements of a comprehensive research programme that includes the drilling of a number of deep boreholes are presented. This programme will be executed over the next five years at an estimated cost of 200 million Swiss francs. It should result in repository projects which guarantee the safe and final disposal of all types of nuclear waste produced in the country. A timetable for the realization of the Swiss concept is included.
1. INTRODUCTION
Nuclear waste from the use of radioactive materials in hospitals, industry and nuclear research has accrued in Switzerland for many decades. Since 1969 the bulk of this has been produced by commercial nuclear plants’. Table I reviews the nuclear situation in Switzerland. Today there are 4 reactors in operation with a total capacity of 1940 MWe.
Low-and medium-level waste from medical institutions, industry and research is collected annually under the supervision of the Federal Office of Health. Conditioning and packaging takes place at the Federal Institute for Reactor Research (EIR) at Würenlingen.
With the exception of combustible materials, the low-and medium-level waste from the operation of nuclear power plants is processed for final storage at the plant site.
93
94 ISSLER et a l
As in Würenl i n g e n , the waste is first reduced in volume by incineration and compression techniques. It is then mixed with cement or bitumen and packed into concrete-lined drums of 200 1 each.
Each power plant has at its site an intermediate storage capacity for 10 years' waste production. Since 1969 Switzerland partic i p a t e d in the annual deep sea disposal campaigns carried out under the control of OECD/NEA w i t h in the framework of the London Convention. In this manner a major portion of Switzerland's low and medium level waste has been disposed of to date.
Under the original reprocessing contracts concluded with France and Great Britain the wastes from spent fuel e l e ments reprocessed before 1979 will remain at these plants. New contracts, signed last year, foresee the option of returning the wastes resulting from the reprocessing of spent fuel elements back to Switzerland. For technical reasons this possibility will not arise before 1990.
In addition to final repositories for low- and medium-level waste, Switzerland will therefore have to provide in the longer term a disposal site for high-level waste within its territory. The operators of nuclear power plants are also working on projects of a central facility for intermediate storage of spent fuel assemblies.
The legal authority for the peaceful use of nuclear en e r gy in Switzerland is laid down in the Atomic Law of 1959, recently revised by Parliament. The revised version c o n tains, among others,generally severe provisions c o n cerning nuclear waste:
the producers of nuclear waste are responsible for its storage and final disposal -,
new nuclear plants can only be built if sufficient d e mand for the new energy can be proved and the final and safe disposal of all wastes is guaranteed;
for n u c l e a r .plants indicated as "planned" in table I the start-up licence is made dependent on 'the existence of a feasible and acceptable project for waste disposal .
TABLE I. NUCLEAR POWER PLANTS IN OPERATION, UNDER
CONSTRUCTION AND PLANNED IN SWITZERLAND
IAEA-SM-243/160 95
Plant/Type Capacity MWe Year of start-up/Status
Beznau PWR 350 1969 in operationBeznau PWR 350 1971 in operationMühleberg BWR 320 1972 in operationGosgen PWR 920 1979 in operationLeibstadt BWR 942 1981 under constructionKaiseraugst BWR 925 1985 plannedGraben BWR 1140 1986 plannedCH - 1 LWR С. 1000 c. 1990 planned
With the passing of the new law the further development of nuclear energy in Switzerland has thus become dependent on fulfilment of the legal request for proof of safe and final disposal of nuclear wastes.
On the other side, the amended legislation has created a special and somewhat easier procedure for obtaining the necessary permits for test drillings and foresees the possibility of land expropriation.
The federal authorities have requested the operators of existing nuclear power stations to prepare by 1985 p r o jects for final repositories.
In anticipation of this request, the producers of nuclear waste already in 1972 formed the National Cooperative for the Storage of Nuclear Waste ,NAGRA, with the aim of clarifying the possiblities of waste disposal. NAGRA "has now been charged by its associates to comply by 1984 with the federal request for repository projects.
TABLE II. WASTE CATEGORIES AND CORRESPONDING WASTE REPOSITORIES IN SWITZERLAND
WASTECATEGORY
KIND OF WASTE ISOLATION TIME REQUIRED
MAIN REQUIREMENTS FOR REPOSITORY
TYPE OF REPOSITORY POSSIBLE GEOLOGIC FORMATIONS
I SCRAP FROM DISMANTLING OF NPP AND VERY LOW ACTIVE WASTES
YEARS TO DECADES
ISOLATION OVER SOME DECADES ; ARTIFICIAL BARRIERS ALLOWED
NEAR SURFACE ROCK CAVERNS
SEVERAL
II MEDIUM- ANDLOW-LEVELWASTES
SOMEHUNDREDYEARS
GEOLOGIC BARRIERS; WITH STABILITY OVER MIN. 500 YEARS
ROCK CAVERN AT 100 TO 600 m DEPTH
ANHYDRITE; CLAY; SPECIAL HYDROLOGIC SITUATIONS
III HIGH-LEVEL WASTES FROM REPROCESSING
SOMETHOUSANDYEARS
GEOLOGIC BARRIERS ; STABILITY OVER MIN. 10 ООО
DEEP BOREHOLES AT 600 m TO 2 500 m DEPTH
GRANITE; BEDDED SAL': ANHYDRITE; CLAY; FORMATIONS WITH STAGNANT WATER
YEARS
IAEA-SM-243/160 97
Parallel to these efforts and in co-ordination with NAGRA the Federal Institute for Reactor Research is conducting a research programme on the problems of nuclear waste d i s posal .
2. THE SWISS CONCEPT OF N U CLEAR WASTE REPOSITORIES
Early in 1978 the Swiss utilities and NAGRA in an elaborate report (Die nukleare Entsorgung in der Schweiz, February1978, VSE Postfach, Zürich)presented their plans and p r o grammes for the final disposal of nuclear waste.
In the following the main features of the Swiss concept as far as they concern the storage aspects will be reviewed:
to arrive at an appropriate final storage scheme the nuclear waste has been grouped into 3 categories.Criteria used for the subdivision were, in addition to radio-activity and toxicity, the technical barriers provided by solidification and. the methods of packaging. The resulting subdivision is explained in greater detail in Table II •
each category of waste has been assigned a corresponding repository which has to satisfy the safety requirements (e.g. length of isolation period) for this category.
3 types of repositories are visualized:
repository type A - for waste category 1 including largevolume components from the m a i n t e nance and decommissioning of nuclear p l a n t s .The necessary isolation time is 100 years in shallow underground caverns.
repository type В - for waste category 2. Required isolation time 1000 years, final storage in underground caverns at a depth less than a few hundred meters in suitable geological formations.
repository type С - for waste category 3. Required isolation time 10 000 years, final storage in suitable geological formations at 600 - 2500 meters depth. Burial in caverns or deep boreholes.
98 ISSLER et al.
3. GEOLOGICAL CHARACTERISTICS OF SWITZERLAND AND THE POTENTIAL HOST ROCKS CONSIDERED
The Swiss portion of the earth's crust is composed of a great variety of rocks. Practically all rock types considered suitable for waste repositories are either outcropping or have been encountered in caverns, galleries and deep b o r e holes. Some of these formations, however, are relatively thin and of limited areal extension. Also, the original rock sequences have been disturbed in many regions by the young alpine orogeny.
For large parts of Switzerland there exist excellent g e o l o gical maps. Geological information from the numerous h i g h way and railroad tunnels particularly in the Alps and the Jura mountains permit in many areas a fairly reliable p r o g n o sis of the geological conditions of the regions above the floors of the principal valleys.
Because there are practically no economically exploitable mineral resources in Switzerland, the deeper subsurface of the country is only poorly explored. A few deep exploration wells for hydrocarbons, potash and mineral water have been drilled in the past, predominantly in the Swiss Plain.They locally give valuable information on the basin fill and allow a first evaluation of potential host rocks.The still very wide grid of deep borings will now have to be complemented by NAGRA with a programme of infill boreholes at carefully selected locations.
The following geological formations are being considered by N AGRA as to their suitability as host rocks for nuclear waste :
3.1 Evaporites
Switzerland has no salt domes in its subsurface. Bedded salt in the Trias formation is rather widespread under the Swiss Plain and the Jura mountains. Where they occur at shallow depth they are being economically exploited and excluded as host rocks. At greater depth they offer possibilities for the construction of a repository of type C- Unfortunately, the average thickness of the salt layers seems to be less than 100 meters, which in a final analysis may prove to be insufficient.
Anhydritic rocks of Triassic age are abundant in S w itzerland. In tectonically little disturbed situations their thickness is of the order of a few tens of meters. In
IAEA-SM-243/160 99
the Alps orogenic movements have led to the tectonic accumulation of lenticular bodies of anhydrite that can be several 100 meters thick.
Anhydrite has many physical and chemical properties which make it attractive as a host rock. To name a few: it is practically impermeable to liquids, possesses excellent rock-mechanical properties and has the interesting ability to seal incipient water-bearing fissures and joints through chemical transformation into gypsum. In addition, the thermal conductivity of anhydrite is very high and its sorption properties are good.
Anhydrites as potential host rocks for waste categories1 and 2 and probably also 3 have been studied in Switzerland for many years. Additional test borings are still r e quired to determine the geometry and physico-chemical p r o perties of some of the geologically more attractive o c c u r rences .
3.2 Argillaceous Formations
Clay and marls are widespread in Switzerland in outcrops as well as in the subsurface to depths of many 1000 meters. They attain thicknesses of several 100 meters particularly in alpine regions where, due to tectonic stresses, they are generally transformed into relatively brittle shales. Preliminary observations in galleries and tunnels indicate that these alpine shales are increasingly plastic and impermeable at greater depth. The suitability of clays, marls and shales as host rocks for waste categories 2 and 3 will be investigated further with in situ experiments and b o r e h o l e s .
3 . 3 Crystalline Rocks
Granites, gneisses and schists form the bed rock u n d e r laying the sedimentary series of the Jura mountains and the Swiss Plain. In the Alps, these igneous and metamorphic rocks are brought to the surface in the Central Massivs.Here they show considerable jointing and fracturing,which strongly impairs their impermeability to percolating grou n d water. Under increasing overburden the ground-water c i r c u lation could well decrease at greater depth below the valley floors. It is assumed that relatively impervious zones exist at depth exceeding 1000 meters. This working hypothesis will be tested by several boreholes at selected l o c a t i o n s .
1 0 0 ISSLER et al.
Granites and gneisses are considered potential host rocks both for waste categories 1 and 2 and, particularly under the Swiss Plain, for category 3.
3.4 Repositories in Geological Formations above the G r o u n d water Level
In the widespread sandstone and limestone formations of Switzerland conditions for the construction of waste r e positories above the regional ground-water level are an attractive possibility. Although limestones and sandstones have much poorer sorption properties than argillaceous sediments, their rock-mechanical behaviour is much better. To be certain of the suitability of a site, the water p e r colating from the surface must be prevented from p e n e t r a ting the host rock by a thick covering layer of clays or marls. Several potentially favourable situations for the disposal of waste categories 1 and 2 have already been identified. They are now being investigated in greater d e t a i l .
3.5 Repositories in Geological Formations containing stagnant Fossil Water___________________________________
If it can be proved that porous formations such as sand- .stones contain water of syndepositional origin, it must follow that in these formations hardly any water c i r c u l a tion has taken place in the geological past. Such f ormations, therefore, are isolated from the biosphere and could serve as host rock for waste category 3.
In the Molasse formations of the Swiss Plain the d e s cribed conditions could exist in certain areas. A com p r e hensive study of well data from exploration tests, supplemented by age determinations of formation waters from additional experimental boreholes will have to verify this c o n c e p t .
4. N A G R A ’S PROGRAMME FOR THE FINAL DISPOSAL OF NUCLEAR WASTE
The time-table for the final disposal of nuclear waste is outlined in Table III.In the medium term the research p r o gramme of NAGRA will have to give proof that in Sw i t z e r land a safe and final disposal of all nuclear waste in accordance with the stipulations of the Atomic Law is p o s s i b l e .
IAEA-SM-243/160 101
TABLE III. TIMETABLE FOR THE FINAL STORAGE OF NUCLEAR WASTE
IN SWITZERLAND
1979 elaboration of disposal concept, determinationof physical and chemical data on waste c onditioning, technical barriers and host rocks
computer programme for safety analysis
preparation of geological drilling programme
1980 procurement of permits for test drillings
1980-84 execution of drilling programmecomplementary field and laboratory investigationspreparation of general non-site-specific repository projects
selection of suitable repository sites safety analyses
application for general construction permits
1985 site-specific projects for final repositoriesfor all categories of nuclear waste which meet the safety requirements under the Atomic Law
additional investigations
1989 Disposal site for waste categories 1 and 2 ready
To this end, tangible repository projects will have to be produced by 1985 demonstrating the following:
suitable repository sites for all waste categories;
engineering projects, plans and procedures for the construction, operation and eventual sealing of the repositories and projects for intermediate storage facilities
102 ISSLER et al.
safety analyses, demonstrating the feasibility and soundness of the proposed schemes.
4.1 Geological Investigations
In the coming five years a number of test borings to a depth of 3000 meters and numerous experimental galleries in potential host rocks are expected to provide the information for the selection of repository sites. These acti v i ties will be supplemented by rock-mechanical measurements, migration studies, heater experiments and experiments on fracture hydrology both in the laboratory and in situ. A research programme to improve the reliability of age d e t e r minations of formation waters is also being prepared.
4.2 Repository Projects
Preliminary system analyses for the construction, operation and sealing of final repositories for all categories of waste are in preparation. These studies also include p r o jects for intermediate storage.
4.3 Safety Analyses
Preliminary safety considerations regarding the problems of isolation, immobilisation and migration of radionuclides have been analysed. Through the refinement of the applied models and the construction of appropriate computer programmes the safety analyses, complemented by the forthcoming input data from the geological research programme, will gradually be improved and concretised.
4.4 Waste Treatment
The Federal Institute for Reactor Research is carrying out a research programme for the conditioning of waste categories 1 and 2. Leaching experiments on solidified waste era-, bedded in cement are being conducted. The corrosion effects of the various packing materials on conditioned waste will also be studied. The institute also investigates the b e haviour of various packaging materials under radiation and the retention capacity (Kd-values) of selected geological f o r m a t i o n s .
IAEA-SM-243/160 103
In view of the still widely held misconceptions on origin, volumes and types of radioactive waste and the near c o m plete ignorance of the protection afforded by a final re pository, NAGRA conducts an agressive public relations c a m paign as an integral part of its research programme. A bi- monthly information bulletin informs the interested public of the progress of its work. Research reports are being . published and scientific discussion of results is e ncouraged through conferences and seminars.
4 . 5 Public Relations
.4 . 6 Cost of Programme
Preliminary estimates put the cost of NAGRA's research p r o gramme for the period up to 1985 at about 200 million Sfr. These expenditures will have to be borne by the producers of nuclear waste. The financing of the programme is already s e cured by the Swiss utilities and the Federal Government. The investments in this programme as well as the construction and maintenance of the storage and repository facilities will increase the cost of electrical energy from nuclear plants by only'a few tenths of a Swiss cent per KWh.
DISCUSSION
K.G. ERIKSSON: If a site is proposed by NAGRA, could it be vetoed by the
local community or be subject to any kind of local referendum?
H. ISSLER: Under the revised Atomic Law the authority to grant permission
for test drillings and site investigations is vested in the Federal Government. There
is no provision for a referendum by the community concerned.
A STRATEGY FOR THE DISPOSAL
OF RADIOACTIVE WASTES IN ITALY
M. MITTEMPERGHER,
CNEN, Dipartimento Radiazioni e
Ricerche di Sicurezza e Protezione,
CSN-Casaccia, Rome, Italy
Presented by W. Bocola
Abstraa
A STRATEGY FOR THE DISPOSAL OF RADIOACTIVE WASTES IN ITALY.The Italian strategy in the field of radioactive waste disposal must face the needs deriving
from present nuclear development and the demonstration of feasibility of technical solutions which are likely to become of importance only in the middle-to-long term, following the evolution of the nuclear industry. The most immediate and real needs are to dispose of medium- to-low activity beta-gamma wastes which are currently kept on provisional storage sites located close to the production centres. For these wastes two dumping operations in the Atlantic Ocean are foreseen in the near future within the NEA programmes. Studies have been undertaken and are actively being pursued in the meantime in order to locate and qualify two sites for the disposal of low- and medium-activity wastes into superficial geological formations. These will become.the final settlement for the disposal of such radioactive wastes. Wastes of high activity or containing transuranic elements at the present rate of production and storage do not require any urgent definitive confinement. However, work has already been in progress in order to ascertain the suitability of Plio-Quatemary clays as geological disposal sites for wastes of long half-life. This work, which is of great importance in order to satisfy public opinion, is carried out in collaboration with the European Communities. It provides for the study of the rates of dispersion o f heat within the geological formation and possible effects thereupon; the characterization of geochemical formation and possible effects thereupon; the characterization of geochemical barriers to radionuclide diffusion; the analysis of models for disposal operations; and the building of an experimental disposal station. The design for disposal within the geological formations will be such as to safeguard the carrying capacity of the clay formation, even though this might mean exluding the possibility of future stored waste recovery. In very general terms, the strategy of disposal will point to the use of geological formations which are common and widespread: this should allow the location of fuel cycle plants in the vicinity of the disposal sites, thus limiting the proliferation of nuclear sites and the transportation problems o f radioactive wastes.
1. INTRODUCTION
The problem of nuclear waste disposal in Italy is increasing in importance and
a solution is urgently needed; this may seem to contrast with the modest and
increasingly difficult development of nuclear activities in Italy.
IAEA-SM-243/67
105
106 MITTEMPERGHER
Currently these developments include four power stations, for a total capacity
of about 1400 MW(e), two reprocessing pilot plants engaged in experimental work,
some installations for UF6 conversion and fabrication of fuel elements for water
reactors and some nuclear centres engaged in research and development activities.
Plants under construction will not affect the short-term situation as regards radioactive
waste production; in fact, these plants are demonstration reactors and power reactors
which will not start operating before 1984. This current slowing down of nuclear
activity in Italy is the result of remarkable difficulties in plant siting and, more
generally, of the difficulties deriving from lack of public acceptance. It is well
known that public opposition to nuclear energy is partly fed by the controversy
on the disposal of radioactive wastes, in particular the long-lived ones. This
controversy is supported by the so-called undemonstrated capability of modem
technology to assure the multimillennial confinement of radionuclides contained
in the wastes and it is also supported by the environmental and geological situation
of our planet.
One of the main aspects of our strategy is to carry out demonstrations and
activities which will reassure public opinion on the reliability of waste disposal
solutions. As a consequence, the strategy itself is aimed to provide a guarantee and
thus obtain public acceptance of nuclear power.
This paper attempts to provide a partial explanation for an evident lack of
balance between programmes of activity and actual achievements.
2. CATEGORIES OF WASTES: PRESENT INVENTORY AND FORECAST OF
FUTURE PRODUCTION
Tables I and II give some indications of the magnitude and type of wastes
produced up to the present time. Table III contains forecasts of future waste
production, as presently envisaged in the most plausible hypotheses of nuclear
development and consequent technical options. From the three Tables, the following
indications can be derived as regards a disposal strategy for the wastes.
(1) Wastes with low and medium beta-gamma activity already represent an urgent
management problem both as concerns the quantities involved and because
they are spread across all the national territory. They are moreover connected
with extremely varied activities, often very scattered, such as those connected
with medical and industrial uses of radioisotopes. Of course, a quantitative
forecast of such wastes over time depends on the effective rate of nuclear
power plant installation. Even if the rate should actually be lower than that
envisaged in Table II, the need to achieve their disposal in practice should not
change.
IAEA-SM-243/67 107
TABLE I. VOLUME OF RADIOACTIVE SOLID WASTES (m3) STORED IN
ITALY UP TO 31 DECEMBER 1978
Source
beta-gamma wastes alpha-wastes
lowactivity
mediumactivity
Electric nuclear plants 1300 1100 -
Fuel cycle pilot plants 1500 300 50
Research centres 2400 1 300
Total 5200 ~ 1400 350
(2) Wastes containing transuranic elements, including those highly active ones
coming from reprocessing operations, are not at the moment arising in such
quantities to require and justify a complete management cycle, including
final disposal. Forecasts of future production are also rather uncertain and
much delayed, being connected with the problem of future reprocessing
options. But despite the absence of short- and medium-term requirements,
the problem of the disposal of such types of waste exists for the following
reasons:
As mentioned above, acceptance of nuclear energy is often connected
with the demonstrated capability to manage its most ‘difficult’ wastes.
Pilot and demonstration activities on the treatment of such wastes
must be supplemented by verification of the feasibility of their disposal.
Lacking an autonomous capability for industrial reprocessing of exhausted
fuel in the short and medium term, it may still be necessary to resort to
reprocessing services abroad. This would imply the necessity of with
drawal of the wastes by the user; then, the problem of their disposal
would arise again.
(3) Tables I—III do not show data concerning plant decommissioning and data on
wastes coming from mining activities for uranium production. Without going
into the subject of decommissioning, which is of extreme importance from
108 MITTEMPERGHER
TABLE II. VOLUME OF LIQUID ALPHA WASTES (m3) STORED IN ITALY
UP TO 31 DECEMBER 1978
Source High-activitywastes
Low-activitywastes
EUREX pilot plant 350 —
ITREC pilot plant 81 -
CASACCIA nuclear centre - 8
the viewpoint of nuclear strategy and plant design, some of the actions
associated with tlus problem will be recalled later on. Not considered here
are the problems of uranium mining tailings. In fact, management of such
materials is associated with technical and protection choices in situ more than
whole-disposal solutions for which a transport stage is unavoidable. In situ
radioprotectional actions must guarantee the absence of air-transported
particulates, must limit leaching activity by circulating waters, and must avoid
superficial stream washing.
3. DISPOSAL OF WASTES WITH LOW BETA AND GAMMA ACTIVITY
Currently, these wastes are stored in provisional deposits in the places of
production. Also, the situation as regards their treatment and conditioning is not
homogeneous and standardized over national territory. The strategy of disposal
of these wastes envisages the following sequence of interventions:
( 1 ) Establishment of a national service with operational and management
duties. In the near future, CNEN, together with industrial utilities, will
establish a company which will take care of the complete and standardized
management of all low beta- and gamma-activity wastes produced in Italy.
The solution of this problem has priority in any national approach to radio
active waste management. The presence of CNEN in this organization is
meant as a warranty of protection of the welfare of the community.
(2) Organization, within the next two years, of a first waste dumping opera
tion in the Atlantic Ocean, in the framework of a specific NEA enterprise,
of a first stock of wastes already treated. Such a first dumping operation
TABLE III. FORECAST OF RADIOACTIVE WASTES (m3) PRODUCED IN ITALY UP TO 2010
SourceYear
1975 1980 1985 1990 1995 2000 2005 2010
Research centres— Beta and gamma
wastes 2400 3600 4800 6000 7200 8400 9600 10 800— Alpha wastes 1600 2400 3200 4600 4800 5600 6400 7200
Open-cycle nuclear power plants a— Beta and gamma
wastes 1300 2900 5300 16 200 37 000 60 000 82 000 103 000— High-activity
alpha wastes (irradiated fuel) 68 171 630 1554 2480 3400 4350
Gosed-cycle nuclear power plants b— Beta and gamma
wastes 1300 2700 4900 14 500 34 00 0 -4 0 000 59 0 0 0 -6 2 000 84 000-85 000 108 000— Alpha wastes - - - - 11 500 25 00 0 -2 7 000 36 000-41 000 52 000— High-activity
alpha wastes - - - - 210 330-490 660-745 950
a 15 MW(e) nuclear power is forecast to be in operation between 1986 and 1990.b The different figures for the years 1995, 2000 and 2005 are determined by the various assumptions on the size and start-up of reprocessing
plant.
IAEA
-SM
-243/6
7
109
1 1 0 MITTEMPERGHER
might be followed, still in the short term, by a second one, for
which the waste treatment will be arranged in advance. With these two
operations to be carried out in the next four years, all low beta- and gamma-
activity wastes should be confined.
(3) For a longer-term strategy, geological solutions for the disposal of these wastes
are envisaged. For such purpose, some geological and hydrogeological superficial
models were defined. Such patterns seem to allow an effective confinement
and a limited and controlled possibility of radionuclides to be leached; these
patterns also correspond to environmental situations widely found in nature.
Areas where such favourable situations are present have been mapped and
among them two sites are being assessed for qualification for national waste
disposal.
(4) For a medium- and long-term solution of geological disposal of short-lived
radioactive wastes and components of plants involved in decommissioning
operations, the possibilities offered by mines already exhausted or being
exhausted are also being assessed. In this case, too, research is aimed at
obtaining an inventory of potentially favourable situations, from which
choices shall be made, involving territorial policy decisions.
4. LONG-LIFE WASTE DISPOSAL, ROLE OF GEOLOGY
The national situation, as indicated above, implies, in terms of operating
needs, medium- and long-term solutions. Nevertheless, the existence of the well-
known psychological problems connected with public acceptance of nuclear energy
imposes in the short term the need for some activities demonstrating the feasibility
of the proposed solutions. For this reason, conceptual and experimental activities
on waste disposal have been formulated over several years and have been co-ordinated
and encouraged by the CEEA.
The approach to the problem of long-lived nuclear wastes has been preceded
by a conceptual analysis of the criteria of multimillennial confinement which must
be met. From such analysis, the opportunity emerged to hypothesize waste
disposal management solutions completely relying on the characteristics of the
natural environmental containment. In other words, this choice ascribed to the
geological environment the function of screening and containment of nuclear wastes.
This means future possibilities of recovering the same wastes are obviously excluded.
The solutions which foresee disposal with the possibility of waste recovery involve
a very different technological approach and a long-term engagement of management.
IAEA-SM-243/67 111
In this case, the geo-environmental conditions would assume the function of a
screening device and, as such, would be the subject of comparative analyses, also
with completely artificial solutions.
To support the completely ‘geological’ solution of multimillennial disposal,
some consideration of the main requirements and of their existence in nature is
necessary: the effectiveness of chemical and geochemical barriers to the radionuclides’
mobility, the geomorphological stability of the area, the technology of construction
of the deposit and of restoration of the original geological conditions.
Chemical and geochemical stability imply complete isolation of the radio
nuclides and their non-interference with the external water cycle.
The study of the mineral deposits demonstrates that anomalous concentrations
of chemical elements with physico-chemical properties similar to those of radio
nuclides, under given geological conditions, are absolutely immobile for long ,
geological periods, that is, for tens of millions of years.
The determining geological conditions to avoid rapid processes of mobilization
and of transport of radionuclides confined in depth are, however, rather generalized
and spread in the Earth’s crust: they are geolithological situations with no active
groundwater circulation and not subject to anomalous thermal flows.
These natural situations may encourage conceptual solutions for the geological
disposal of long-lived wastes which are valid in terms of chemical and geochemical
isolation, taking also into account the low geochemical mobility of most actinides,
that is, of the most toxic elements.
The geomorphological stability refers to the requirement that a certain deep
radioactive waste deposit, owing to a rapid geological evolution of the area, does
not move to a more superficial situation and, consequently, does not become
involved in hydrological and erosive processes that would lead to its dispersion.
The present state of knowledge of Earth Sciences allows for each area an
exact and detailed reconstruction of its geological history, over tens of millions
of years.
Modem sophisticated geochronology techniques using natural isotopes,
palaeontology, sedimentology and palaeomagnetism, permit a clear reconstruction
of the palaeogeographical, tectonic and morphological evolution for each environ
mental situation. It is clear that a careful choice of the area for a deposit will
imply such a reconstruction and that perfect historical knowledge of the last
million years permits safe forecast extrapolations for the next 100 000 years.
Based on such extrapolations, it should be possible to determine the optimum
depth of disposal.
The problem of the available technologies for the deep storage of nuclear
wastes and for the subsequent restoration of the conditions proper to the selected
geological body has, in fact, already been solved in the present state of mining
mechanization.
112 MITTEMPERGHER
Without discussing in detail the models of possible deposits, or the specialist
mining techniques, it is sufficient, as an example, to recall the sophisticated
technologies of modern oil drilling which allow the reaching of depths of over
10 km and operation on very deep marine beds. These refined methods of
approach and operation, widely in use in oil research and exploitation, therefore,
can guarantee the possibility of geological disposal of long-lived radioactive wastes
with multimillennial containment.
5. RESEARCH ON PLIOCENE AND QUATERNARY CLAYS
IN SOUTHERN ITALY
Definitive geological confinement of long-lived radioactive wastes implies a
conceptual approach which differs greatly from that of geological disposal with the
possibility of waste recovery.
Apart from the common requirements of seismotectonic and geomorphological
stability, the definitive geological solution implies the existence of very precise and
strict natural conditions which, in the alternative solution, may, at least in part,
be created artificially. One may refer in particular to hydrological and hydro-
geological conditions, to the thermal impact characteristics, and to the validity of
chemical, geochemical and physico-chemical barriers with respect to radionuclide
migration. As a consequence, "should the hypothesis of definitive geological disposal
be pursued, the choice of a suitable geological formation and the studies for its
characterization will require an important research effort in many branches of the
Earth Sciences.
Since 1968, the prospect of definitive geological disposal in Italy led to the
choice of Neogene clays as the possible natural multimillennial geological body of
containment. These clays have the following important characteristics:
(1) They are spread over a wide area, along almost all of the peninsula and in
Sicily. Not only is this a valid measurement for an optimum choice, but it
can also favour a logistical connection between the reprocessing plant and the
waste disposal site.
(2) Clays have a high degree of plasticity, which allows them to absorb mechanical
stresses, and gives rise to self-sealing processes in the case of shearing fractures
and drilling. This represents a significant containment guarantee in a country
such as Italy with active tectonic movements.
(3) Generally, clay deposits have a physico-chemical character with negative
redox values, which determine reducing characteristics, while the pH almost
always has basic values. These two characteristics, together with the ionic
IAEA-SM-243/67 113
exchange capacity that many clay minerals have both with respect to
transuranic elements and to the long-lived fission products, should eliminate
or greatly limit radionuclide migration.
Laboratories, studies and field experiments carried out up to the present
time on clays have followed three different lines:
(1) Geological evaluation to identify the optimum stratigraphie and tectonic
conditions for a future disposal.
(2) Detailed laboratory study of the natural characteristics and on-site tests in a
sample area, in order to evaluate a first set of values such as the natural
chemical and physico-chemical barriers and heating impact.
(3) On-site and laboratory studies on milling technologies applied to clay
formations.
Geological and geographical evaluations are made in subsequent steps: since
we are dealing with Neogene clays the geological reference unit is the sedimentary
basin. Such approaches involve evaluation of the sedimentary, stratigraphie,
mineralogical, geochemical and hydrological parameters; palaeographic and
neotectonic reconstructions give indications about the probable dynamic evolution
of the basin. Judgement on the possible use of a basin also entails verifying eventual
mining interests which may directly involve the clay formation. In this connection,
salt deposits and gas pockets intercalated within the clay levels are also important.
Laboratory tests and field experience in sample areas allow an even more
refined knowledge of mineralogical, hydrological and geochemical parameters that
are relevant to possible radionuclide migration, both under natural conditions and
during thermodynamically anomalous transients, such as artificially induced clay
heating by the wastes.
A third cycle of experiments, which has started only recently, concerns
research and testing of mining systems and technologies suitable for location of
wastes in the lithological context and for the restoration of the original isolation
conditions of the geological medium.
For the time being, such experiments are concentrating on drilling technologies,
on various aspects of mining materials and technologies used for temporary hole
covering and the possibility of recovering casings. The natural subjects of such
experiments are the geotechnical characteristics of clays at different stages of
consolidation and the parameters of the plastic response of such materials to
mining interventions.
It is expected that the results of the three cycles of experiments will provide
preliminary answers as regards feasibility. When more conclusive results are
available, it should be possible to start the conceptual design of disposal technologies
in clays and, finally, the design and construction of a first pilot waste disposal
11.4 MITTEMPERGHER
centre. The development of research on the three groups of problems mentioned
above will probably take at least five years. Later on, when research is nearly
completed, the design and pilot construction stages should logically follow but
they will require the effective availability of a site.
REFERENCES
[1] BOCOLA, W., MITTEMPERGHER, М., Considerazioni sulla gestione dei rifiuti radioattivi, Ingegneria Nucleare 6, November-December 1977, p. 3 (in Italian).
[2] GERA, F., MITTEMPERGHER, М., “ Gestione dei rifiuti radioattivi” , Simposio sul Ciclo del Combustibile Nucleare, ANDIN, Pugnochiuso, 1978, Ingegneria Nucleare 3, May M ay-June 1978, p. 3 (in Italian).
[3] NEA-OECD, Objectives, Concepts and Strategies for the Management of Radioactive Waste Arising from Nuclear Power Programmes, Report by a NEA Group of Experts,September 1977.
IAEA-SM-243/128
LES ACTIVITES DE RECHERCHE
ET DE DEVELOPPEMENT DES
COMMUNAUTES EUROPEENNES EN MATIERE
D’EVACUATION DES DECHETS RADIOACTIFS
DANS LES FORMATIONS GEOLOGIQUES
P. VENET, E. Delia LOGGIA, W. FALKE,
B. HAIJTINK, Ph. MASURE
Commission des Communautés européennes,
Bruxelles
Abstract-Résumé
THE EUROPEAN COMMUNITIES’ RESEARCH AND DEVELOPMENT ACTIVITIES RELATIVE TO THE DISPOSAL OF RADIOACTIVE WASTES INTO GEOLOGICAL FORMATIONS.
The European Communities’ research and development activities in radioactive waste disposal are part of its more general multiyear programmes on radioactive waste management and storage. The immediate purpose of these activities is to determine the best conditions for disposal of high-level and/or long-lived wastes into geological formations so that they do not present any danger to man and his environment. The studies are carried out either under contract with various organizations and firms of Member States on a cost-sharing basis or directly at the facilities of the Joint Research Centre at Ispra. The Communities’ programme at present embraces most of the activities of the Nine in Europe on waste disposal in deep geological formations. There is a co-ordinated division of the work among the national organizations responsible for waste disposal, and this arrangement takes into account existing national commitments to specific formations (such as salt in the Federal Republic of Germany), the particular nature of the subsoil in the territories concerned (such as clay in
Belgium) and considerations of economy to avoid duplication of costly research. Thus, argillaceous formations are being studied at present mainly by Belgium and Italy, crystalline rocks (granite) by France and the United Kingdom and salt domes by the Federal Republic of Germany and the Netherlands. Back-up studies applicable to all these different formations are being conducted by Denmark, Ireland and the Communities’ Joint Research Centre at Ispra. The paper describes the activities and studies being carried out under the Communities’ programme on the various formations concerned, indicates the progress achieved and surveys the results obtained.
LES ACTIVITES DE RECHERCHE ET DE DEVELOPPEMENT DES COMMUNAUTES EUROPEENNES EN MATIERE D’EVACUATION DES DECHETS RADIOACTIFS DANS LES FORMATIONS GEOLOGIQUES.
Les activités des Communautés européennes en matière d’évacuation des déchets radioactifs s ’inscrivent dans le cadre plus général des programmes pluriannuels de la CCE sur la gestion et le stockage des déchets radioactifs. Leur objectif immédiat est de déterminer les meilleures conditions d’évacuation des déchets de haute activité et/ou de longue période dans les formations
115
116 VENET et al.
géologiques, afin qu’ils ne puissent nuire aux hommes et à leur environnement. Les études sont conduites, soit dans le cadre de contrats à frais partagés avec différents organismes et firmes des Etats Membres, soit directement dans les installations du Centre commun de recherche à Ispra. Le programme communautaire regroupe actuellement la majeure partie des travaux effectués dans l’Europe des Neuf dans le domaine de l’évacuation dans des formations géologiques profondes. Une répartition coordonnée des travaux entre les organismes nationaux chargés de l’évacuation des déchets a été effectuée. Cette coordination a tenu compte d’engagements nationaux déjà existants pour une formation spécifique (comme le sel dans le cas de la République fédérale d’Allemagne), de la nature particulière du sous-sol des territoires concernés (comme l’argile dans le cas de la Belgique) et d’un souci d’économie en évitant la duplication de recherches coûteuses. Il en résulte qu’actuellement les formations argileuses sont étudiées principalement par la Belgique et l’Italie, les roches cristallines (granite) par la France et le Royaume-Uni et les dômes salins par la République fédérale d’Allemagne et les Pays-Bas. Des études de soutien, valables pour tous ces types de formations, sont effectuées par le Danemark, l’Irlande et le Centre commun de recherches de la CCE à Ispra. Le mémoire présente, pour les différentes formations concernées, l’ensemble des travaux et études entrepris dans le cadre du programme communautaire, indique leur état d’avancement et donne un aperçu des résultats obtenus.
1. INTRODUCTION
1.1. Les activités des Communautés européennes en matière de stockage et
d’évacuation dans les formations géologiques sont menées dans le cadre plus
général des programmes pluriannuels de recherche et de développement de la CCE
relatifs à la gestion et au stockage des déchets radioactifs. Ces programmes, d’une
durée de 3 à 5 ans, se succèdent depuis 1973 et devraient se poursuivre au cours
des années 80.
Leur objectif est de disposer en temps opportun, dans la Communauté, de
solutions efficaces en vue d’assurer la sécurité des populations et la protection de
l’environnement contre les risques potentiels associés à la gestion et au stockage des
déchets radioactifs. L’action communautaire doit être considérée comme un
complément et un soutien coordonné aux efforts consentis par chacun des Etats
Membres pour réaliser des sites opérationnels de stockage et d’évacuation des
déchets radioactifs de haute activité et/ou de longue période, conçus de telle sorte
que soit exclu le retour de la radioactivité dans la biosphère en quantités
constituant un risque biologique.
1.2. Le programme d’action directe a débuté en 1973. Les études, exécutées dans
les propres laboratoires de la Commission, essentiellement au Centre commun de
recherches d’Ispra (CCR), concernent principalement la séparation et la trans
mutation des actinides ainsi que l’évaluation des risques à long terme présentés par
l’évacuation de ces déchets (voir IAEA-SM-243/161, dans les présents comptes
rendus).
IAEA-SM-243/128 117
Le programme d’action indirecte a commencé en 1975; les travaux sont
conduits par les laboratoires nationaux, dans le cadre de contrats à financement
partagé passés par la Commission avec des organismes publics ou privés des Etats
Membres. Les études concernent le traitement des déchets radioactifs et l’évacua
tion des déchets de haute activité et/ou de longue période dans les formations
géologiques ainsi que les aspects administratifs, juridiques et financiers liés à leur
gestion.
Ces programmes se termineront à la fin de 1979. Bien qu’aucune décision
formelle n’ait encore été prise à ce sujet par le Conseil des Ministres des pays de
la Communauté, le programme d’action directe devrait être prolongé jusqu’en
1983 et le programme d’action indirecte continué jusqu’en 1984. Le premier sera
centré sur les sujets suivants: évaluation des risques, barrières de protection,
séparation et contrôle des actinides. Quant au futur programme d’action indirecte,
il prolongera le précédent en l’élargissant à l’évacuation dans les dépôts géologiques
sous-marins et à l’enfouissement à faible profondeur.
Le tableau I précise le déroulement du programme communautaire et les
budgets totaux des actions directe et indirecte. On constatera que les dépenses
annuelles totales des programmes communautaires sont passées d’environ 2,0 MUC1
par an en 1973 à 10,4 MUC par an en 1975; elles devraient atteindre 16,2 MUCE
par an à partir de 1980. Le budget total du premier programme d’action indirecte
relatif à l’évacuation dans les formations géologiques s’est monté à environ 29 MUC
dont 13 MUC ont été financés par la CCE.
1.3. Une répartition coordonnée des travaux entre les organismes chargés de
l’évacuation des déchets a pu être réalisée. Cette coordination a tenu compte
d’engagements nationaux déjà existants (par exemple celui concernant le sel en
Répub lique fédérale d ’Allemagne), de la nature spécifique du sous-sol des territoires
nationaux concernés (l’argile en Belgique) et d’un souci d’économie en évitant
des duplications inutiles. Il en résulte qu’actuellement les formations argileuses
sont étudiées principalement par la Belgique et l’Itaüe, les roches cristallines
(granité) par la France et le Royaume-Uni et les dômes salins par la République
fédérale d’Allemagne et les Pays-Bas. Des études de soutien, valables pour tous ces
types de formations, sont effectuées par le Danemark, l’Irlande et le Centre commun
de recherches de la CCE (étabUssement d’Ispra).
On notera cependant qu’actuellement ces choix n’entraînent aucunement, à
l’exception de la République fédérale d’Allemagne, un engagement définitif pour
le type de formations étudiées.
1 1 MUC (million d’unités de compte) = 1 MUCE (million d’unités de compteeuropéennes) = 1,3 X 106 dollars des Etats-Unis.
TABLEAU I. DEROULEMENT DU PROGRAMME COMMUNAUTAIRE ET BUDGETS DES ACTIONS DIRECTE
ET INDIRECTE EN MATIERE D’EVACUATION DE DECHETS RADIOACTIFS DANS LES FORMATIONS
GEOLOGIQUES
Programmes et budgets8
Années
73 74 75 76 77 78 79 80 81 82 83 84
Action directe
(CCR)b
D1
6,1 MUC
D 2
21,1 MUC
D 3
22,2 MUCE
Action indirecte
I 1
28,9 MUC
12
58,9 MUCE
Budget total4,1 MUC 52 MUC 81,1 MUCE
5 X 106 dollars 68 X 106 dollars 105 X 106 dollars
8 L’unité de référence est l’UC (unité de compte) qui équivaut à environ 1,3 dollars des Etats-Unis; depuis le 1er janvier 1978 cette unité
a été remplacée par l ’UCE (unité de compte européenne); 1 MUC représente un million d’unités de compte.
b CCR: Centre commun de recherches, Ispra.
IAEA-SM-243/128 119
Les activités scientifiques et techniques concernant l’évacuation des déchets
radioactifs de haute activité et/ou de longue période dans des formations géologiques
continentales comportent trois phases principales.
La première phase, de conception générale et d ’évaluation de la sécurité à
long terme, est celle qui a bénéficié jusqu’ici de l’effort scientifique le plus important
dans le cadre du programme communautaire. Cette phase s’attache en premier
lieu à la définition des principes généraux de l’évacuation dans les formations
géologiques et à la conception des installations de dépôt. Elle concerne en outre
la recherche et la caractérisation des sites géologiques potentiellement aptes à
recevoir des déchets radioactifs. Dans ce cadre, des méthodes d’analyse et des
techniques de reconnaissance et de mesure nouvelles sont conçues de manière à
fournir une image relativement précise des équilibres internes et à long terme des
formations géologiques. Ces études ont déjà débouché sur l’élaboration d’une
première série de modèles mathématiques représentatifs. L’analyse des potentialités
de migration des radionucléides jusqu’à la biosphère se base par ailleurs sur des
études de géoprospective et sur l’élaboration de scénarios probabilistes d’évolution
du site à long terme.
L’ensemble de ces travaux concourt à la détermination des effets à long
terme de l’évacuation des déchets radioactifs sur la biosphère et des risques qui y
sont liés.
La deuxième phase, ou phase expérimentale de dépôt et d ’analyse semi-
déterministe de l ’efficacité des barrières naturelles et artificielles à long terme, vise
d’abord à la connaissance détaillée du massif en profondeur à partir de puits et
galeries exécutés à l’emplacement de zones représentatives des formations
sélectionnées. C’est ainsi,que des essais et mesures sont actuellement réalisés in
situ dans le sel (RFA) et le seront bientôt dans l’argile (Belgique). Si cette phase
d’observation est jugée positive, la décision de réaliser des installations d’évacuation
expérimentales, comprenant le contrôle et l’auscultation détaillés de l’environnement
naturel et artificiel, pourrait être prise par les Etats Membres concernés. Ces
installations pourront faire partie, au plan technique, du programme de la
Commission, mais resteront à ce stade sous la responsabilité de chaque Etat. Il
sera ainsi possible de mettre au point et d’éprouver in situ les techniques de
construction et d’exploitation les mieux adaptées.
L’observation de l’évolution des équilibres internes du miüeu soumis au
dégagement thermique et aux rayonnements permettra par ailleurs de préciser les
modèles mathématiques et de les adapter à la structure interne réelle du site
d’évacuation. L’analyse semi-déterministe permettra à ce stade de faire une
évaluation précise de l’efficacité des barrières naturelles et artificielles à long terme.
Elle constituera un argument scientifique essentiel lors d’une prise de décision
concernant la réalisation de l’installation d’évacuation définitive.
2. CADRE SCIENTIFIQUE ET TECHNIQUE
TABLEAU II. ETAT D’AVANCEMENT DU PROGRAMME COMMUNAUTAIRE POUR L’EVACUATION DES DECHETS
RADIOACTIFS DANS DES FORMATIONS GEOLOGIQUES
Granite Argüe Sel Etudes générales
Actions CEA UKAEA CEN/SCK CNEN G SF ECN NRC AEC CCRIGS K FK Ris<ÿ
France Royaume-Uni Belgique Italie R FA Pays-Bas Danemark Irlande
Sélection des sites * * * * * * * *
Caractérisation des sites- sondages profonds - + * + * -
- hydrogéologie + + + + +- propriétés physiques, chimiques et
géomécaniques de la roche + + + + +- propriétés thermiques du milieu géologique + + + + +
Confinement et rétention des radioéléments- conditionnement + + + = +- barrières artificielles et m atériaux de
remplissage des cavités + = + =
Evolution des barrières naturelles etartificielles après dépôt- impact du dégagement thermique + ’ + + + +- interaction déchets-environnement
géologique et non géologique + + + + +
Modélisation de la migration des radioéléments et analyse des risques associés + + + + + + +
Conception des installations d’évacuation + + + + + +
Réalisation de chambres expérimentales = *
1979 1974
= programmé + en cours * terminé
120 VENET
et al.
IAEA-SM-243/128 121
La phase ultime, ou phase opérationnelle d ’évacuation définitive, s’attachera
au projet détaillé des installations de dépôt, à leur réalisation et à la gestion
complète de l’exploitation jusqu’à sa fermeture définitive. Elle comportera
également la conception, la mise en place et l’exploitation d’un réseau de contrôle
détaillé et de surveillance à long terme du milieu géologique et de son environnement.
En tout état de cause, cette phase ne sera pas abordée au cours du prochain
programme 1980—1984.
3. ETAT D’AVANCEMENT DES PROGRAMMES COMMUNAUTAIRES
Les études entreprises dans le cadre du programme ont été regroupées en
quelques grands thèmes (tableau II). On se reportera, pour connaître le détail de
certaines de ces activités et les résultats obtenus, aux mémoires présentés à ce
colloque par plusieurs des laboratoires et organismes ayant participé au programme
communautaire.
3.1. Sélection des sites
Le premier pas en vue de cette sélection a été l’établissement d’un catalogue
européen des formations géologiques présentant des caractéristiques favorables à
l’évacuation des déchets radioactifs solidifiés de haute activité et/ou de longue
période. Ce catalogue a été établi grâce à la collaboration permanente des instituts
de géologie des Etats Membres et sur la base de critères d’ordre géologique, choisis
en commun.
Une carte européenne à 1:1 500000 résume cette sélection pour les formations
argileuses, salines et cristallines situées sur les territoires des Etats Membres, à
l’exception du Luxembourg, de la Corse et du Groenland ainsi que de l’île de Man,
des îles anglo-normandes et des îles Feroe. Une version condensée du catalogue
devrait être disponible en 1980.
S’appuyant sur cette étude, un certain nombre de sites particulièrement
représentatifs des formations géologiques retenues ont été sélectionnés dans
plusieurs pays membres afin d’analyser in situ leurs caractéristiques géologiques
et de connaître les conditions de confinement réelles qu’elles pourraient présenter
dans le cadre de l’évacuation des déchets radioactifs.
3.2. Caractérisation des sites
La caractérisation des sites sélectionnés dans chacune des formations
géologiques étudiées est actuellement à un stade d’avancement variable suivant les
pays. Elle est encore incomplète, l’effort le plus grand devant porter sur la connais
sance précise des paramètres caractérisant l’hydraulique souterraine, la répartition
des contraintes naturelles, les propriétés thermiques et la structure profonde des
massifs.
122 VENET et al.
Pour les formations cristallines, les reconnaissances détaillées ont porté sur
huit sites représentant une grande variété de massifs granitiques au Royaume-Uni
et sur deux sites appartenant à des provinces géologiques différentes en France (ces
deux sites ont été retenus après une étude systématique assez détaillée de nombreux
autres). Dans l’ensemble, les forages de reconnaissance ont été superficiels et
l’exécution de forages profonds (500 à 1000 m) ne devrait commencer que dans
les prochains mois. Les études de surface mises en oeuvre sont généralement très
complètes: cartographie géologique, analyse structurale (télédétection, photo-
interprétation, statistique des discontinuités), néotectonique et sismicité, évolution
des lignes de rivage, techniques géophysiques (sismique réfraction, prospection
électrique et électromagnétique, gravimétrie), etc. Les études hydrogéologiques
combinant des essais in situ à faible profondeur et des mesures de surface ont permis
de faire un premier bilan hydrogéologique à l’échelle des massifs (France et
Royaume-Uni) et d’examiner les conditions de migration de traceurs chimiques en
subsurface (France) (IAEA-SM-243/68).
Seuls les essais hydrauliques qui pourront être exécutés lors de la réalisation
des sondages de grande profondeur permettront de préciser la fiabilité du confine
ment à long terme dans les massifs granitiques. Des mesures de conductivité
thermique du granite ont été faites in situ en Comouailles (Royaume-Uni),
parallèlement à un programme détaillé de mesures de laboratoire sur carottes de
sondage. Par ailleurs, une méthode sophistiquée de datation des eaux (Royaume-
Uni) basée sur la mesure du déséquilibre 234U/230Th et 234U/238U devrait permettre .
de définir des zones de faible circulation des eaux souterraines.
Pour les dépôts argileux, la Belgique a concentré ses reconnaissances sur
l’argile de Boom rencontrée dans le sous-sol du Centre d’étude de l’énergie nucléaire
de Mol (IAEA-SM-243/2). Une campagne de sondages profonds ainsi qu’un
programme de reconnaissances géophysiques par sismique réfraction ont permis de
connaître en détail la structure géologique du site, ainsi que les conditions hydro
géologiques régnant dans les formations encaissantes, plus perméables. Les
caractéristiques minéralogiques, physiques et mécaniques de l’argile ont été déter
minées au laboratoire, à partir d’échantillons non remaniés prélevés in situ. Des
analyses chimiques et radiochimiques des eaux souterraines ont également été
faites. Des essais in situ sont en cours pour analyser les impacts minéralogique,
physique, hydraulique et mécanique du dégagement thermique sur les argiles de
Boom, dans la carrière de Terhagen.
De la même façon, la formation argileuse rencontrée à Trisaia, en Italie, sur
le site du CNEN, a pu être caractérisée grâce à la mise en oeuvre d’un programme
de reconnaissances géophysiques complété par une campagne de reconnaissances
par sondages. Un programme de forages profonds et d’essais thermiques in situ
commencé l’an dernier a été interrompu récemment à la suite de difficultés
techniques.
IAEA-SM-243/128 123
Le dôme salin qui a donné lieu aux reconnaissances les plus complètes est
celui de la mine désaffectée de sel gemme (halite) de Asse (RFA). Compte tenu
du fait que la structure géologique du site a été définie très précisément tout au
long de la période d’exploitation minière, les travaux de reconnaissance ont surtout
porté sur la caractérisation des propriétés minéralogiques, physiques et géo
mécaniques de la matrice rocheuse ainsi que sur l’analyse détaillée des effets du
dégagement thermique sur le comportement rhéologique du sel (IAEA-SM-243/15).
La détermination des contraintes naturelles dans le massif est également en cours.
3.3. R étention des radionucléides
Les propriétés de rétention naturelle des ions transportés par l’eau susceptible
de circuler dans les formations géologiques sont d’une importance majeure lorsque
l’on analyse les potentialités de migration des radionucléides depuis le site d’évacua
tion jusqu’à la biosphère. Aussi, un important travail de détermination en
laboratoire des caractéristiques de sorption-désorption s’est-il développé dans les
divers pays de la Communauté pour les trois types de formations géologiques
étudiés. Les coefficients de distribution K<j ont été déterminés pour divers produits
de fission (Sr, Cs, Eu, I) et actinides (diverses formes de Pu, Np, Am, Ra).
On a analysé leur variation avec la température, le pH des solutions et la
concentration en ions. Toutefois, un gros effort méthodologique reste à faire pour
obtenir des résultats d’essais représentatifs des phénomènes réels, en particulier
dans les milieux à porosité de fissure comme le granite. A ce sujet, un essai
d’injection de traceurs a été réalisé in situ en France, dans un milieu granitique, ce
qui a permis de faire des observations sur le transfert des ions beaucoup plus
réalistes que la simple extrapolation des essais de laboratoire.
Une autre méthode de mesure in situ a été mise au point au Danemark, par
injection de traceurs dans la formation géologique à partir d’un forage, puis
pompage avec détermination des profils de concentration en traceurs restitués.
Parallèlement à ces mesures de transfert dans les formations géologiques, une
recherche systématique de matériaux susceptibles de constituer une barrière géo
chimique de rétention efficace pour les radionucléides libérés par une éventuelle
lixiviation des produits de conditionnement a été faite, notamment en France, où
l’on a pu mettre en évidence la grande capacité de rétention de l’attapulgite, l’illite
et la bentonite (IAEA-SM-243/155).
3.4. Evolution des barrières naturelles e t artificielles après dépôt
La connaissance de l’évolution à moyen et long terme des barrières naturelles
et artificielles susceptibles de constituer le confinement des déchets radioactifs
est d’une importance primordiale.
124 VENET et ál.
Les conteneurs de déchets vitrifiés et les revêtements des cavités d’évacuation
peuvent être considérés comme une barrière artificielle de durée de vie limitée. Une
série de travaux réalisés en Belgique, en France, en RFA et au Royaume-Uni a
permis de faire une présélection des matériaux susceptibles d’être utilisés pour leur
fabrication. Des essais de corrosion ont débuté récemment et il est encore pré
maturé de tirer des conclusions concernant le choix des matériaux, la structure des
conteneurs ou leur géométrie.
Par ailleurs, une action de recherche intégrée entre la France, la RFA et le
Royaume-Uni a été organisée pour évaluer l’évolution à long terme des verres de
conditionnement (dévitrification, effets des rayonnements alpha, etc.) et leur
résistance à la lixiviation par des solutions aqueuses provenant du milieu géologique.
L’impact le plus immédiat et le plus sensible de l’évacuation des déchets radio
actifs de haute activité sur l’environnement géologique sera lié à leur dégagement
thermique. Celui-ci aura des effets très variables suivant la nature des formations
géologiques réceptrices.
Pour les formations granitiques, des méthodes de calcul analytique ont été
développées au Royaume-Uni. Elles permettent d’évaluer l’influence du dégagement
thermique sur la répartition des contraintes à différentes époques dans un massif
théorique considéré comme semi-infini, continu, homogène et isotrope. On a ainsi
estimé les variations de perméabilité qui pourraient en résulter (IAEA-SM-243/26).
En France, un programme d’études similaire est en cours, fondé sur un modèle
de calcul par éléments finis qui permettra d’analyser l’influence des variations de
contraintes d’origine thermique sur la stabilité des ouvrages souterrains pendant
les phases d’exploitation du site d’évacuation. Il permettra également l’examen
systématique des modifications que l’on pourrait apporter aux principales
caractéristiques du dépôt et leur influence sur la réalisation technique et sur les
coûts du projet global. Cette étude de sensibilité porte en particulier sur la
profondeur du dépôt, sa capacité, la puissance thermique des déchets évacués, la
géométrie des conteneurs, la mise en place de barrières physico-chimiques artifi
cielles, la réversibilité du dépôt et la réfrigération éventuelle du site d’évacuation.
Pour les dépôts argileux, de nombreux essais de laboratoire ont été réalisés
en Belgique, au Danemark et en Italie, sur des échantillons non remaniés de
dimensions variables (10 cm à 80 cm) afin de mesurer la conductivité thermique
de l’argile, mais également pour connaître les modifications de ses caractéristiques
minéralogiques, physiques et géomécaniques et donc de son comportement en
fonction de la température. Ces essais et mesures effectués en laboratoire ou en
carrière (site de Terhagen, Belgique), seront complétés ultérieurement par des essais
réalisés dans une chambre expérimentale équipée, à 220 m de profondeur, sur le
site de Mol (Belgique).
■ Pour les formations salines, de nombreux essais in situ ont été effectués en
RFA, dans la mine de Asse II, afin de définir la réponse des formations de sel
gemme (halite) à un dégagement thermique élevé. Un programme de mesures de
IAEA-SM-243/128 125
convergence des parois des galeries et trous de forage a servi de base à l’élaboration
d’un modèle mathématique tenant compte du comportement rhéologique observé
in situ. Parallèlement, des essais sur échantillons de laboratoire soumis à des
températures constantes ont permis d’analyser les caractéristiques de diffusion de
l’eau interstitielle contenue dans le sel. Une confirmation des résultats obtenus
doit être fournie par un essai d’échauffement in situ d’une durée de 6 mois
effectué à 775 m de profondeur dans la mine de Asse. Enfin, on a mis en évidence
l’existence de risques de migration de gouttelettes de saumure contenues dans la
formation saline vers l’intérieur du trou de forage, en cas de forte élévation de
température engendrée par le dépôt de déchets de haute activité. Enfin, l’évolution
prévisible des températures dans les dômes salins après dépôt de déchets de haute
activité a été calculée aussi bien en RFA qu’aux Pays-Bas (IAEA-SM-243/104).
Un programme de synthèse des connaissances relatives à l’impact de la
température et des rayonnements portant sur le milieu granitique, mais également
sur les formations argileuses et salines, est actuellement développé (Belgique, France,
RFA). Il traite plus particulièrement de l’évolution des caractéristiques texturales
et structurales (rupture des inclusions de fluide par exemple), minéralogiques
(altération hydrothermale), géochimiques (lixiviation, radiolyse, diffusion inter
cristalline), géomécaniques (résistance et comportement rhéologique, expansion
thermique) et hydrauliques (perméabilité, porosité cinématique) du milieu.
3.5. Modélisation des écoulements en milieu de faible perméabilité et estimation
des risques liés à la migration des radionucléides jusqu’à la biosphère
La complexité des phénomènes intervenant dans le transport des radio
nucléides et la difficulté de caractériser le milieu géologique ont jusqu’ici imposé
l’approche mathématique pour l’étude de la migration des isotopes radioactifs
libérés lors de la lix iviation des déchets conditionnés.
Les premiers modèles développés s’appuient tous sur les équations de diffusion
de Fick.
En Belgique, un modèle tridimensionnel aux éléments finis a été mis au point
afin d’analyser l’influence relative des divers phénomènes physico-chimiques qui
affectent la migration des radioéléments dans le milieu argileux poreux saturé .
Le programme élaboré permet de tenir compte de la convection et de la dispersion
des éléments dans la phase liquide, mais également des phénomènes de précipitation,
d’adsorption, de rétention et de décroissance radioactive.
En France, le modèle tridimensionnel aux éléments finis mis au point pour
analyser la migration des radioéléments dans le granité utilise les mêmes équations
de diffusion de Fick et fait donc appel au concept de milieu poreux équivalent au
milieu fissuré (IAEA-SM-243/68). Il est composé de deux modules, le premier
représentant un modèle d’écoulement hydraulique, le deuxième simulant les
phénomènes de transfert proprement dit en tenant compte de la convection, la
dispersion, la rétention ionique (sorption-désorption) et la décroissance radioactive.
126 VENET et al.
Ce modèle a été utilisé pour interpréter des essais d’injection de traceurs à 40 m
de profondeur dans un massif granitique. Il est actuellement testé pour prendre
en compte les effets du dégagement thermique sur la répartition des contraintes
(donc des perméabilités) dans le massif et sur le gradient hydraulique induit.
Au Royaume-Uni, les études de migration des radioéléments dans le milieu
granitique discontinu s’appuient sur une équation monodimensionnelle (programme
FACSIMILE).
Enfin, dans le cas des formations salines, les Pays-Bas ont mis au point un
modèle adapté aux conditions géologiques particulières rencontrées dans la région
étudiée (où les dômes salins sont protégés par des dépôts argileux superficiels) en
estimant les concentrations de radioéléments libérés dans la biosphère dans
l’hypothèse de remontée lente des dômes salins par diapirisme. Il est certain que
tous ces modèles devront être adaptés aux futures observations faites in situ, lors
des prochaines phases de reconnaissance, visant à une connaissance plus précise et
plus fidèle des caractéristiques du milieu géologique et des phénomènes d’écoulement
des fluides en milieux de faible perméabilité.
L’analyse de la probabilité d ’occurrence d ’événements géologiques susceptibles
d’affecter l’intégrité de la barrière géologique a été développée tant au Centre
commun de recherche d’Ispra (IAEA-SM-243/161) que dans plusieurs pays de la
Communauté à partir d’une analyse d’arbres de défauts basée sur l’identification
d’événements primaires ou de séquences d’événements naturels ou induits susceptibles
de se produire sur un site déterminé et pendant des périodes de temps fixées (phase
opérationnelle, 1000 ans, 25 000 ans, 100 000 ans par exemple). Si l’approche
méthodologique peut être considérée dès à présent comme élaborée, il n’en va pas
de même pour l’information scientifique disponible sur les événements géologiques
naturels et artificiels envisagés, qui est limitée (IAEA-SM-243/25, IAEA-SM-243/68).
Après l’utilisation des modèles de migration des radionucléides au sein du
milieu géologique, l’estimation des effets de la dissémination éventuelle des radio-
isotopes dans la biosphère (nappe phréatique, surface du sol et atmosphère) constitue
le dernier élément de l’analyse du risque. Jusqu’ici, une première série d’estimations
sur les quantités de radionucléides parvenant à la surface dans le cadre de scénarios
de migration sélectionnés a été réalisée (Royaume-Uni, Norvège, Belgique, RFA,
France).
3.6. Conception et réalisation d’installations expérimentales d’évacuation
La réalisation de chambres expérimentales en profondeur constitue la base
la plus solide pour connaître de manière détaillée la structure et les propriétés
internes du massif, pour analyser les déséquilibres occasionnés par l’excavation et
le dégagement thermique sur le régime hydraulique et l’état de contraintes pré
existant, pour préciser les conditions techniques de réalisation des installations
souterraines et les éprouver.
IAEA-SM-243/128 127
Le fonçage d’un puits d’accès à une chambre expérimentale en milieu argileux
à 220 m de profondeur sur le site de Mol (Belgique) commencera à la fin de
1979. La mise en oeuvre d’un important programme d’essais et mesures in situ
permettra de connaître avec précision le comportement et les caractéristiques
géotechniques, thermiques et hydrauliques de la formation argileuse. L’exécution
même du puits nécessitera probablement la congélation des terrains plastiques pour
assurer la stabilité et l’étanchéité de cet ouvrage qui traversera des formations
perméables contenant un aquifère situé au-dessus des dépôts argileux. Elle
permettra de préciser les techniques de réalisation d’un ouvrage définitif les plus
appropriées. La possibilité d’utiliser un tunnelier pour l’excavation de galeries
horizontales, qui se heurte au problème de convergence des parois par fluage de
l’argile, sera également analysée.
Des expériences en formations salines en vue du stockage à titre expérimental
de déchets de haute activité (RFA) ont pu être développées dans les galeries de la
mine de Asse, exploitée par ailleurs au plan national pour le stockage de déchets
de basse activité.
En dehors des études détaillées concernant la caractérisation du site et l’impact
du dégagement thermique sur la formation (halite) qui ont été présentées dans les
chapitres précédents, le programme expérimental a porté sur les conditions pratiques
d’évacuation. En particulier, on a pu mettre au point des techniques de manutention
et de transport souterrain des conteneurs jusqu’à leur dépôt dans des trous de forage,
ainsi qu’un système de récupération postérieure éventuelle. Un élément de fermeture
provisoire des trous de forage a été également étudié pour la phase d’exploitation.
La fermeture de la mine de fer de Konrad (IAEA-SM-243/14), toujours en
RFA, a donné l’occasion d’étendre le concept de dépôt dans le sel à d’autres forma
tions géologiques profondes. Un programme d’études est actuellement en cours,
non seulement pour analyser la fiabilité du confinement par le milieu géologique
(colithes coraliennes ferrifères), mais également pour définir les modifications qu’il
serait nécessaire d’apporter aux installations souterraines existantes pour utiliser
cette mine comme éventuel site d’évacuation.
Les études de conception des installations d ’évacuation sont restées à un stade
très général dans la plupart des pays de la Communauté. Elles consistent pour le
moment à comparer divers projets de configuration, du point de vue de la sécurité,
de la fiabilité, des techniques d’exécution et des coûts. Les projets élaborés,
actuellement très proches les uns des autres, paraissent dès à présent techniquement
réalisables (IAEA-SM-243/3, IAEA-SM-243/93).
4. CONCLUSION
Le programme communautaire a permis de regrouper dans un ensemble
cohérent la majeure partie des travaux effectués, tant dans les Etats Membres que
128 VENET et al.
dans les propres laboratoires de la Commission, en matière d’évacuation dans les
formations géologiques profondes de déchets radioactifs de haute activité et/ou
longue période.
Il a également contribué à l ’établissement d ’un véritable dialogue entre les
organismes de recherche chargés de l’évacuation des déchets. Des groupes de
travail ont été constitués et se réunissent régulièrement, de même que sont
organisés des séminaires et des réunions techniques sur des sujets très spécifiques.
Ces réunions à caractère restreint, ainsi que la communication de nombreux
rapports sur l’état d’avancement des travaux assurent l’information détaillée et
actualisée des participants sur les études en cours.
Il y a lieu d’ajouter que le soutien de la Commission a également permis
d’entreprendre ou d’accélérer des recherches qui seraient restées sans cela à un
stade embryonnaire au niveau national.
Au plan technique, l ’inventaire des formations géologiques de la Communauté
potentiellement favorables à l’évacuation des déchets radioactifs a fait apparaître
que des choix nombreux et variés s’offrent aux autorités responsables pour la
prospection puis l’étude d’éventuels sites d’évacuation.
On peut dès à présent affirmer que la réalisation d ’installations d ’évacuation
est possible en faisant appel aux techniques actuelles, quelles que soient les
formations géologiques concernées (sel, argile, granite) mais moyennant le respect
d’un certain nombre de conditions que le programme a largement contribué à
mettre en évidence.
Toutefois, il n’apparaît pas possible aujourd’hui de prétendre qu ’un type de
formation présente une plus grande aptitude qu’un autre à recevoir des déchets
radioactifs; à la limite il pourrait même s’avérer que chaque type de formation a
des mérites particuliers vis-à-vis de catégories différentes de déchets.
Dans le même ordre d’idées, d’autres formations présentant des potentialités
intéressantes pour l’évacuation des déchets radioactifs pourront également entrer
dans le cadre des études du prochain programme communautaire.
En dernier lieu, les connaissances acquises permettent désormais de définir
avec précision les actions à entreprendre et les points sur lesquels les efforts devront
se focaliser, à savoir:
— les interactions entre le milieu géologique et les déchets conditionnés;
— l’impact du dégagement thermique sur l’environnement géologique;
— l’évaluation de l’efficacité du confinement à long terme dans les conditions
d’évacuation envisagées.
Ces actions nécessiteront la réalisation de nombreuses études intégrées
regroupant des spécialistes d’activités très diverses et exigeront une coordination
accrue au niveau communautaire.
Il est clair que l’évacuation des déchets radioactifs est une matière d’intérêt
commun à de nombreux pays, comme le montre abondamment le présent colloque.
IAEA-SM-243/128 129
Pour sa part la Commission des Communautés européennes s’est efforcée, dans un
premier stade, d’ouvrir ses programmes à une coopération élargie avec l’aide d’autres
organisations internationales. Cette coopération s’est surtout manifestée par
l’organisation conjointe de séminaires et de réunions techniques avec la collaboration
de FAIEA et de l’Agence de l’énergie nucléaire de l’OCDE. La Commission pour
suivra ses efforts dans ce domaine.
C’est dans cet esprit que sera organisée par la Commission, du 20 au
23 mai 1980 à Luxembourg, la première Conférence de la Communauté européenne
sur la gestion et le stockage des déchets radioactifs, où seront présentés à la
communauté scientifique internationale les résultats des activités de recherche.
IAEA-SM-243/99
Invited Paper
SUBMARINE GEOLOGIC DISPOSAL OF
NUCLEAR WASTE
C.D. HOLLISTER, B.H. CORLISS
Department of Geology and Geophysics,
Woods Hole Oceanographic Institution,
Woods Hole, Massachusetts
D.R. ANDERSON
Sandia Laboratories,
Albuquerque, New Mexico,
United States of America
Abstract
SUBMARINE GEOLOGIC DISPOSAL OF NUCLEAR WASTE.
Site suitability characteristics of submarine geological formations for the disposal of
radioactive wastes include the distribution coefficient of the host medium, permeability,
viscoelastic nature of the sediments, influence of organic material on remobilization, and
effects of thermal stress. The submarine geological formation that appears to best satisfy
these criteria is abyssal “red” clay. Regions in the ocean that have coarse-grained deposits,
high or variable thermal conductivity, high organic carbon content, and sediment thickness
of less than 50 m are not being considered at this time. The optimum geological environment
should be tranquil and have environmental predictability over a minimum of 105 years. Site
selection activities for the North Atlantic and North Pacific are reviewed and future activities
which include international cooperation are discussed. A paleoenvironmental model for
Cenozoic sedimentation in the central North Pacific is presented based on studies of a long
core from the Mid-Plate Gyre MPG-1 area, and is an example of the type of study that will be
carried out in other seabed study areas. The data show that the MPG-1 region has been an
area of slow, continuous accumulation during the past 65 million years.
GEOLOGIC SITING CONSIDERATIONS FOR THE DISPOSAL OF RADIOACTIVE WASTE INTO SUBMARINE GEOLOGIC FORMATIONS
Before selecting a generic submarine repository study site for any radioactive waste one must complete the exercise of establishing, and then ranking, site suitability conditions.
In this preliminary effort, to identify submarine geologic formations that appear most suitable in an environment of maximum stability the following media and site characteristics appear desirable:
131
132 HOLLISTER et al.
I . THE GEOLOGIC MEDIUM I S THE PRIM E BA RRIER TO R ELEA SE
It is the fundamental working criterion that the "host medium" or geologic medium remains the single most important barrier to the release of radioactive material to the biosphere. This broad generalization leads to the two most important retention aspects of the host medium— high sorption capability and low permeability.
k. The distribution (or sorption) coefficient (Kd) of the host medium is a measure of its ion uptake capacity. The distribution coefficient is simply the ratio of the amount of material (radioactive ions in this case) that is bound to the medium to the amount that is free to travel in the pore water through the medium either by diffusion or advection. A Kd of 1 implies that for every ion locked up one is mobile; a Kd of 10° says that the ratio is a million locked up to one that is mobile. Kd values are related to grain size amongst other parameters, i.e. the finer-grained sediments have the highest Kd's. A high Kd for a broad range of elements is one of the key desirable characteristics of a disposal medium.
B. Another important characteristic is permeability, the measure of the speed with which fluids can migrate through a medium. For waste retention, low rates of flow are desirable. Permeability should not be confused with porosity, which is simply the measure of pore space in a medium. The important point is that a medium such as a fine-grained clay may be very porous but have a very low permeability because the pores are not connected. A medium (such as silt or sand) may have a comparatively low porosity but be highly permeable if its fabric includes a large fraction of interconnecting pore spaces.
C. Another desirable characteristic of the medium is that it should behave plastically or as a viscoelastic medium to enhance its integrity if disturbed. Such plasticity could also produce self-healing of an emplacement hole.
D. The role of organic material in processes of remobilization is unknown due to the difficulty of modeling the effect of organics in complexing ions and thus preventing their sorption by the sediment. Our initial studies have focused on sediments with the least amount of organic material.
E. Another even less well understood (or quantifiable) characteristic of the medium is its ability to remain undisturbed under high or variable thermal stress. That is to say, it is necessary to understand the medium's behavior under a variety of thermal fields in order to predict its response as a
IAEA-SM-243/99 133
function of temperature and time. A fundamental question here is whether the existence of a heat source in the repository could cause the sediment to convect and thereby breech the sedimentary barrier.
F. In summary, the most desirable characteristics of the "host medium" are:
1. low permeability and high Kd,2. ability to self-heal, i.e. be visco-elastic in response to
dynamic stress,3. stability under predicted thermal loading,4. a low content of organic matter, i.e. be well oxidized.
The submarine geologic formation that appears to best satisfy the above criteria is abyssal 'red' clay. Depending on organic interactions and permeability considerations, light brown deep-sea clays with 20-40% CaCOg also may be suitable. Together these clay deposits cover over one half of the deep sea floor. Increasingly organic-rich, more permeable biogenic oozes appear less suitable, with turbidite sands and silts least desirable of all.
Thus for the time being the U.S. Seabed Program is not considering areas known to contain geologic formations with the following characteristics:
(a) Coarse-grained deposits such as the proximal portions of abyssal plains (including all fracture zone floors), and ponded turbidite sands of intermountain valleys that are expected to have high permeabilities and low Kd's and are least likely to behave plastically. This criterion effectively eliminates all submarine canyons, continental shelves, shallow portions of the Mid-Oceanic Ridge and fracture zone floors.
(b) Deposits exhibiting high or variable thermal conductivity or where heat flow gradients are non-linear. These regions lack coherency of thermal properties,which probably reflects variations in geotechnical properties, most probably bulk permeability. It is too early in the program to review all heat flow data in order to exclude areas on a global scale; rather it seems prudent to do the exclusion exercise on a larger scale with other more general exclusion criteria and then to review heat-flow data and perhaps conduct special heat flow cruises in candidate regions.
(c) Deposits with high organic carbon content;-in order to avoid the nonpredictable behavior of chelated radioactive isotopes. This criterion would probably eliminate the continental
134 HOLLISTER et al.
rise and slope where the organic carbon content of the upper tens of meters of hemipelagic material ranges between 1 and 10%.
(d) Sediment thickness of less than 50 meters. From initial calculations and from experience with the free fall Giant Piston Corer it appears that a penetrometer will reach between 20 and 40 meters depth below bottom in the types of sediment that fit the above selection criteria. Because of the high and variable bulk permeabilities that have been determined for at least the upper layers of oceanic crustal basalt a buried container should be well clear of basement.
II. THE OPTIMUM GEOLOGIC ENVIRONMENT FOR THE MEDIUM SHOULD BE TRANQUIL AND PREDICTABLE.
This aspect of the assessment, i.e. setting or environment as contrasted with medium characteristics, also yields constraints based on certain working assumptions. The most important of these is environmental predictability over the minimum time span needed for waste isolation (i.e. 10^ years). By this it is meant that the goal is to be able to predict, with high levels of confidence, the probability of a natural event occurring at a repository that would alter, in a detrimental way, its suitability as a repository. The initial approach has been,and still is, to identify areas of low geologic activity,i.e. geologically tranquil regions with ver.y low seismic and volcanic (tectonic) activity and a long history of continuous deposition. Such regions are found in the centers of the large lithospheric plates.
Thus one of our site exclusion criteria is proximity to the edges of lithospheric plates. This excludes those areas within 100 miles of any recorded seismic event greater than magnitude xl or within 100 miles of volcanoes that are known to have been active over the past 10? years. All deep-sea trenches, mid-ocean rift valleys, the crests of the Mid-Oceanic Ridge, and zones of transform faulting are excluded from consideration.
Another geologic perturbation that affects the earth is recurring ice ages. Present trends suggest that another ice age will occur within the next 5 000 to 50 000 years. This is less than the 10^ year isolation period specified above and thus regions likely to be adversely affected by another ice age should be excluded from consideration. These areas include all areas known to have experienced ice-age erosion and all deposits containing coarse, ice-rafted debris.
1 This value will be established after completion of our geotechnical/dynamic
response program.
IAEA-SM-243/99 135
Because eventually man will seek any natural resource of value, all regions containing known or suspected concentrations of food, minerals or hydrocarbons should be excluded. This includes regions where sand or gravel can be or may be mined (i.e. all continental shelves). From the point of future hydrocarbon recovery the entire continental margin including deep sea fans, cones and aprons should also be excluded.
In addition, regions below areas of high biological productivity, such as upwelling areas, which support major fisheries, are excluded and regions of manganese nodules rich in nickel and copper are also excluded. Inasmuch as these sur- ficial deposits could be mined in advance of a disposal operation, this criterion might be relaxed in the future.
Regions crossed or occupied by (1) telecommunication cables,(2) major shipping lanes, or (3) defense installations are excluded.
In addition to the above considerations, there are certain low level waste criteria governed by the IAEA recommendations (IAEA INF CIRC/205/Add. 1/Rev. 1, August 1978) that should be taken into consideration. Most of these are covered above by the high-level waste criteria, but they are listed here for completeness.
1. Sites should lie between 50° North and 50° Southlatitude^
This criterion is designed to avoid sources of bottom water which are characterized by strong vertical mixing, and areas of high biological productivity in the polar regions. Ice rafting is also a concern in such areas.
2. Depth at the site should be 4000 meters' or more.This criterion is derived from the fact that biological, chemical, physical and topographical gradients generally decrease below 4000 m; bottom water circulation is slower; and organic carbon in the pelagic sediments of such areas tends to be low. This criterion is also motivated by a desire to be clear of continental margins.
3. Sites should be remote from continental margins.This 'criterion is designed to avoid regions of high biological productivity, active resource exploration and exploitation, and geologic unpredictability and instability (continental slope, rise and associated fans and canyons).
4. Sites should be away from areas of potential seabed
136 HOLLISTER et al.
resources.This criterion is designed to minimize the likelihood of future disturbances which might shorten pathways to man, and to avoid possible conflicts in "land" uses.
5. Sites should be away from transoceanic cables in use.This criterion is intended to avoid disturbances and conflicts in uses.
6. Sites should be away from areas where geologic hazards such as submarine slides, volcanoes and earthquakes decrease a site's environmental predictability.
This criterion is designed to reduce the likelihood of unpredicted disturbances which might shorten pathways to man.
7. The area of a site should be defined by precise coordinates, with an area as small as practicable.
This criterion appears to be motivated by a desire to limit the affected area.
.8. If possible, sites should be in areas covered by precisenavigational aids.
This criterion is intended to assist in relocating the site for monitoring purposes.
9. Sites should be away from areas, such as submarine canyons, which may unpredictably affect rates of exchange of deep waters and organisms with surface waters near the continental shelf.
This criterion is intended to avoid shortening of the pathways to man.
10. Sites should be chosen for convenient conduct of operations and to avoid, so far as possible, the risk of collision with other traffic and undue navigational difficulties.
This criterion is specified to minimize hazards to navigation and safe operations at the site.
III. SUMMARY OF OCEAN REGIONS EXCLUDED (AT THIS TIME) FROM CONSIDERATION AS A SUBBOTTOM RADIOACTIVE WASTE REPOSITORY?2
1. the continental margin including fans, deltas, aprons, cones,2. proximal portions of abyssal plains,3. all fracture zone abyssal plains,4. all submarine canyon-levee systems,
2 Ranking is not implied at this point in this preliminary siting effort.
IAEA-SM-243/99 137
5. areas covered with less than 50 meters of sediment,6. areas less than 100 nautical miles from plate boundaries,7. areas with ice-rafted debris,8. major shipping lanes, cable routes and defense installa
tions ,9. seafloor regions below areas of high biological produc
tivity.
Approximately one third of the world's ocean floor satisfy these criteria.
A REVIEW OF SITE SELECTION ACTIVITES
The U.S. Seabed Program is focused on the philosophy of bringing a number of candidate sites in the North Atlantic and North Pacific forward in a parallel mode. As we further define the siting criteria we will begin to focus more effort on smaller regions. Our aim is to have identified roughly half a dozen areas on the order of 10^ to 10^ km^ in 1979 and then to conduct some confirmatory cruises with the research vessels of the United States and the research vessels DISCOVERY, RESOLUTION AND CHARCOT during 1980 and 1981.
I. NORTH ATLANTIC
Our first task in selecting any of these study sites for radioactive waste disposal was to reach consensus among leading experts, in open national and international forum, about the above-mentioned suitability criteria.
By applying these criteria to some North Atlantic historical data four large areas were identified, two east of Bermuda in the Western Atlantic Basin (MPG-3 N and S)^, one on the Greater Antilles Outer Ridge^ and one in the Eastern Atlantic Basin^.
All available historical data for these areas are now being collated, plotted and reproduced for distribution to the International Seabed Working Group and for confirmatory cruise planning purposes.
3 MPG = Mid-Plate Gyre
MPG-3 N - 29-34°N, 52°-58°W, 3 X 10s km2
MPG-3 S - 22-29°N, 58°-69°W, 8 X 10s km2
4 Antilles - 20-22°N, 64°-68°W, 8 X 104 km2
5 E. Atl. - 15-32°N, 25°-35°W, 1.7 X 106 km2
138 HOLLISTER et al.
Two cruises (RESOLUTION and ENDEAVOR) into the far northwestern corner of MPG-3 N were conducted in 1978 and the results of this effort suggest that the western part of the MPG-3 N region is very active with respect to current erosion around the perimeter; however, a small area of about 2 500 km^ was identified that might satisfy most, or all, of the criteria if such an area warrants further study.
Assessment of the Greater Antilles Outer Ridge is only in the earliest stage of evaluation and a report on it will be finished in 1980.
The Eastern Atlantic study site was chosen largely on the basis of the site-suitability criteria listed above. A significant amount of historical data and some analysis of unpublished data indicates that within this large area there may be at least two sub-regions that would be worthy of further study: one south and west of the Great Meteor Seamount Chainand the other north and west of the Cape Verde Islands.
Tracks and station data from research vessels have been plotted at our working scale of 4" per degree of longitude. Thetracks and stations have been annotated as to data type andformat and these sheets have been distributed to participatingnations at the Seabed Working Group Meeting in Albuquerque(March 5-7, 1979).
II. NORTH PACIFIC
Most of the initial site selection work for this area has been presented in Sandia progress reports since 1974. The final document concerning the Giant Core obtained in the first MPG-1 region is in the 1978 Sandia Progress Report. Briefly it states that the MPG-1 has been an area of slow, steady accumulation, with no perturbation that would alter its suitability as a repository, during the past 65 million years. There is no data to suggest that the area will undergo any environmental changes over the next million years that would make it unsuitable as a repository. The only reservation may be that the total sediment thickness above bedrock is occasionally less than 50 meters. Areas to the west on older crust, where sediment is thicker, are going to be examined as part of our 1980 effort.
A PALEOENVIRONMENTAL MODEL FOR CENOZOIC SEDIMENTATION IN THE CENTRAL NORTH PACIFIC
Our recent work on a single long core from MPG-1 in the North Pacific provides an example of the type of study that will be conducted in other seabed study areas.
IAEA-SM-243/99 139
A paleoenvironmental model of Cenozoic sedimentation in the central North Pacific has been constructed from sediraento- logical, geotechnical and stratigraphie data derived from a Giant Piston Core (GPC-3: 30°N, 157°W; 5705 m ) . This corerepresents a record of continuous sedimentation for about 65 million years. The core was taken from a region of abyssal hill topography located beneath the present-day carbonate compensation depth. It contains 24.5 meters of undisturbed sediment composed of oxidized brown clay with altered ash layers. Paleomagnetic stratigraphy for the upper 4.5 meters indicates sedimentation rates are 2.5 mm/1000 years for the last2 m.y.6 and 1.1 mm/1000 years before that to 2.4 m.y. Ichthyolith stratigraphy shows relatively continuous sedimentation rates of0.2-0.3 mm/1000 years from 65 to 5 ш.у. Grains greater than 38 micrometers (less than 1 to 2% of the sediment) deposited between 65 and 22 m.y. consist primarily of fish debris.Smectite, phillipsite, feldspars, and clinoptilolite predominate in the less than 20 micrometer fraction. From 22rru y. to the present, the greater than 38 micrometer fraction (less than 1% of the sediment) consists largely of manganese micronodules, whereas the less than 20 micrometer fraction consists of quartz, smectite, chlorite, cristobalite, kaolinite, illite, feldspars and mica.
The observed sedimentological variations can be explained in terms of present and past sedimentation patterns in the central North P a c if ic and by the NNW motion of the Pacific plate during the Cenozoic. Geotechnical properties do not vary raonotonically with depth, due to downcore mineralogical variations in the sediment.
The lack of large hiatuses in GPC-3 indicates that this region has been relatively stable during the Cenozoic in comparison to other regions in the Pacific or other oceans where hiatuses of greater than 10 m.y. duration are known to exist in the sediment record.
6 m.y. = million years.
IAEA-SM-243/112
ПОВЕРХНОСТНЫЕ ХРАНИЛИЩА
ОСТЕКЛОВАННЫХ ВЫСОКОАКТИВНЫХ ОТХОДОВ
А.Н.КОНДРАТЬЕВ, В.В.КУЛИЧЕНКО, И.И.КРЮКОВ,Н.В.КРЫЛОВА, В.И.ПАРАМОШКИН, М.В.СТРАХОВ Государственный комитет по использованию атомной энергии СССР,Москва,Союз Советских Социалистических Республик
Abstract- Аннотация
NEAR-SURFACE STORAGE FACILITIES FOR VITRIFIED HIGH-LEVEL WASTES.
Concurrently with the development of methods for solidifying liquid radioactive wastes,
reliable and safe methods for the storage and disposal of solidified wastes are being devised
in the USSR and other countries. One of the main factors affecting the choice of storage
conditions for solidified wastes originating from the vitrification of high-level liquid wastes
from fuel reprocessing plants is the problem of removing the heat produced by radioactive
decay. In order to prevent the temperature of solidified wastes from exceeding the maximum
permissible level for the material concerned, it is necessary to limit either the capacity of
waste containers or the specific heat release of the wastes themselves. In order that disposal
of high-level wastes in geological formations should be reliable and economic, solidified wastes
undergo interim storage in near-surface storage facilities with engineered cooling systems.
The paper demonstrates the relative influences of specific heat release, of the maximum
permissible storage temperature for vitrified wastes and of the methods chosen for cooling
wastes in order for the dimensions of waste containers to be reduced to the extent required.
The effect of concentrating wastes to a given level in the vitrification process on the cost of
storage in different types of storage facility is also examined. Calculations were performed
for the amount of vitrified wastes produced by a reprocessing plant with a capacity of five
tonnes of uranium per 24 hours. Fuel elements from reactors of the water-cooled, water
moderated type are sent for reprocessing after having been held for about two years. The
dimensions of the storage facility are calculated on the assumption that it will take five years
to fill. It is also assumed that the degree of concentration of liquid wastes at the time of
vitrification and, consequently, the specific heat release of the vitrified materials entering
the store will vary greatly. The specific heat release for solidified materials lies between
5 X 103 and 10s W/m3. The volume of the materials that can be stored varies in inverse
proportion to the specific heat release, i.e. between 9400 and 470 m3. The total heat release
in a full vitrified waste storage facility is 20 MW.
ПОВЕРХНОСТНЫЕ ХРАНИЛИЩА ОСТЕКЛОВАННЫХ ВЫСОКОАКТИВНЫХ ОТХОДОВ.
Одновременно с развитием методов отверждения жидких радиоактивных отходов в СССР и других странах проводится разработка методов надежного и безопасного хранения и захоронения отвержденных отходов. При выборе условий хранения отвержденных отходов, которые получаются в результате остекловывания жидких высокоактивных отходов, образующихся на заводах по переработке твэлов, особое место занимает проблема отвода тепла радиоактивного распада. Чтобы не допустить повышения температуры отвержденных отходов выше допустимого
141
142 КОНДРАТЬЕВ и др.
для данного материала уровня, приходится ограничивать определяющие размеры емкостей для отходов или удельное тепловыделение последних. С целью обеспечения надежного и экономичного захоронения высокоактивных отходов в геологические формации предусматривается предварительное хранение отвержденных отходов в наземных хранилищах с организованным теплоотводом. В докладе показано влияние удельного тепловыделения и допустимой температуры хранения остеклованных отходов, а также выбранных методов их охлаждения на степень необходимого сокращения определяющих размеров емкостей для отходов. Кроме того, рассматривается влияние степени концентрирования жидких отходов в процессе остекловывания на экономичность их хранения в хранилищах различного типа. Расчеты производились для остеклованных отходов, образующихся на перерабатывающем заводе производительностью :по урану 5 т/сут. Твэлы реактора типа ВВЭР поступают на переработку после выдержки около двух лет. Объем хранилища рассчитан на заполнение в течение пяти лет. Принято, что степень концентрирования жидких отходов при остекловывании, а следовательно, и удельное тепловыделение стеклообразных продуктов, поступающих в хранилище, меняются в широком диапазоне. Удельное тепловыделение отвержденных продуктов лежит в пределах от 5-103 Вт/м* до 10s Вт/мэ . Объем хранящегося продукта обратно пропорционален удельному тепловыделению и составляет от 9400 м3 до 470 м3. Общее тепловыделение в полностью заполненном хранилище остеклованных отходов равно 20 МВт.
1. ДОПУСТИМЫЕ РАЗМЕРЫ ЕМКОСТЕЙ ДЛЯ ОСТЕКЛОВАННЫХ ОТХОДОВС РАЗЛИЧНЫМИ ФИЗИЧЕСКИМИ СВОЙСТВАМИ
Допустимые размеры емкостей с тепловыделяющими продуктами определяются условиями теплоотвода, а также удельным тепловыделением и допустимой температурой саморазогрева материалов, выбранных для включения в нихлродуктов деления. На рис. 1 представлены результаты расчета допустимых размеров (диаметров *) емкостей с остеклованными отходами при различных условиях теплоотвода. Хороший теплосъем обеспечивается при применении в качестве теплоносителя проточной охлаждающей воды. В этом случае размеры емкостей могут быть близки к максимальным (кривая 1). Однако остеклованные отходы в хранилищах с водяным охлаждением должны долгие годы находиться под постоянным контролем, так как химическая стойкость даже лучших образцов стекла такова, что необходима их строгая изоляция от воды в течение десятков лет. Применение герметичных пеналов из нержавеющей стали для размещения емкостей с остеклованными отходами увеличивает надежность хранения, но значительно сокращает допустимые размеры емкостей (кривая 2) из-за ухудшения теплоотвода вследствие наличия воздушной прослойки между емкостями и пеналом.
Охлаждение тепловыделяющих продуктов воздушным потоком также приводит к уменьшению размеров емкости. При непосредственном контакте охлаждающего воздуха с поверхностью емкостей это сокращение несколько меньше, чем при введении пенала и наличии водяного охлаждения (кривая 3) , однако в этом случае создаются условия для радиационно-химического разрушения остеклованного продукта,
* Поскольку для цилиндрических емкостей с высотой более четырех диаметров определяющим размером является диаметр емкости, в дальнейшем речь будет идти о допустимых диаметрах емкостей с тепловыделяющим продуктом.
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Рис.1. Зависимость допустимого диаметра емкости от удельного тепловыделения и термической
устойчивости остеклованных отходов.
Водяное охлаждение, температура стекол емкости или пенала равна 1000 С: 1 - непосредствен
ное охлаждение емкостей с остеклованными отходами; 2 — охлаждение через стенку пенала, со
держащего емкости с остеклованными отходами; 6 - непосредственное охлаждение емкостей с
остеклованными отходами.
Воздушное охлаждение, температура воздуха на входе и выходе соответственно равна 40 °С
и 150° С: 3 - при контакте с поверхностью емкости; 4 - через стенку охлаждающей трубы;
5 - при контакте с поверхностью емкости.
что может потребовать организации очистки охлаждающего воздуха от радиоактивной пыли. Чтобы избежать необходимости устанавливать фильтры, емкость с остеклованными отходами может быть отделена от воздушного потока ограждающей трубой. Естественно, что введение дополнительных тепловых сопротивлений на пути теплового потока (к ним относятся материал трубы и воздушный зазор) еще больше сокращает допустимые размеры емкостей. Иногда это может оказаться нежелательным с технологической точки зрения.
Кривые 1- 4, представленные на рис. 1, были получены для остеклованных отходов с максимальной допустимой температурой хранения 700° С и температурой поверхности не выше 400°С . Для сравнения на рис. 1 приведены также кривые, отражающие зависимость от удельного тепловыделения допустимых размеров емкостей для остеклованных отходов с максимальной допустимой температурой хранения 1000°С. Кривая 5 получена для воздушного охлаждения с непосредственным контактом теплоносителя и емкости. Кривая 6 относится к водяному охлаждению с непосредственным контактом теплоносителя и емкости.
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А Б В Г Д Е Ж
Рис.2 . Разрез хранилища остеклованных отходов с воздушным охлаждением: 1 — вентиляторы;
2 — фильтровальная станция; 3 — кран с дистанционным управлением; 4 — разборное перекрытие;
5 - емкости со стекломассой; 6 - каналы для удаления воздуха из хранилища; 7 - перегрузочная
шахта; 8 — транспортер; 9 - каналы для подачи воздуха в хранилище; 10 — железобетонные
колодцы.
Более высокая допустимая температура хранения позволяет увеличить размеры емкости при определенном удельном тепловыделении или дает возможность повысить удельное тепловыделение остеклованных отходов при определенном размере емкости.
2. КОНСТРУКЦИИ ХРАНИЛИЩ ОСТЕКЛОВАННЫХ ОТХОДОВ
Разработка конструкций хранилищ остеклованных отходов проводилась с учетом следующих отправных положений:
1) хранилище размещается на поверхности земли выше уровня грунтовых вод;2) хранилище примыкает к зданию для остекловывания отходов и связано с
ним транспортным коридором;
3) доставка емкостей с остеклованными отходами из здания для остекловывания и размещение их в хранилище осуществляется с помощью дистанционно управляемого крана грузоподъемностью 5 т;
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1 2 3 4 5 6 7 8 9 10 11 12 13 14
UA
Pu с. 3. План хранилища остеклованных отходов с воздушным охлаждением: 1 — перегрузочная
шахта; 2 — отсеки для стеклоблоков; 3 — бетонные колодцы; 4 - фильтровальная станция;
5 — вентиляторы.
4) остеклованные отходы поступают на хранение в емкостях, сделанных из жаропрочной стали марки Х28;
5) приточно-вытяжная система примыкает непосредственно к хранилищу;6) по мере уменьшения количества выделяемого тепла поочередно по отсекам
производится перевод хранилища остеклованных отходов на режим работы станции захоронения отходов, однако сохраняется возможность извлечения емкостей из хранилища.
2.1. Хранилище с воздушным охлаждением
Хранилище состоит из пяти бетонных отсеков (рис. 2 и 3) , внутри которых на определенном расстоянии друг от друга, образуя решетку, расположены железобетонные трубы (колодцы) для размещения емкостей с остеклованными отходами. Каждый отсек рассчитан на заполнение в течение года. Загрузка емкостей в колодцы производится через прямоугольные люки в разборном перекрытии, закрываемые бетонными пробками.
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Рис. 4. Разрез водного бассейна для хранения остеклованных отходов: 1 - кран; 2 - пенал;
3 — бассейн; 4 — перегрузочная тележка; 5 — транспортер; 6 — емкости со стекломассой.
Рассматривались две схемы подвода воздуха для охлаждения емкостей. По первой схеме воздух продувается через кольцевой зазор между емкостями и внутренней поверхностью колодца. Подвод воздуха осуществляется по каналам, проложенным в полу хранилища. После прохождения по колодцам нагретый до 150°С воздух собирается в верхних каналах, проходящих над колодцами, и после очистки выбрасывается в атмосферу. В период загрузки емкостей незаполненные колодцы отключаются от системы путем установки съемной заглушки на канале нагретого воздуха.
По второй схеме охлаждающий воздух омывает наружные поверхности ограждающих труб. Подача воздуха в отсеки и удаление воздуха из отсеков производится по самостоятельным каналам.
Циркуляция воздуха в первый период работы хранилища и в том и в другом случае осуществляется за счет работы вентиляторов. После спада тепловыделения отходов возможен переход на естественную конвекцию (использование вытяжной трубы) . Такой переход может быть осуществлен быстрее при использовании второй схемы, так как в этом случае нагретый воздух не подвергается очистке на фильтрах.
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1 2 3 4 5 6 7 8 9 10
Рис.5. План водного бассейна для хранения остеклованных отходов: 1 - подача пеналов; 2 - пе
регрузочная ишхта; 3 - бассейн; 4 - перегрузочная тележка.
2.2. Хранилище с водяным охлаждением
Хранилище представляет собой железобетонное сооружение (рис.4 и 5) , облицованное внутри нержавеющей сталью и заполненное водой (конденсатом) . Емкости с остеклованными отходами помещаются в герметичные пеналы из нержавеющей стали. Пеналы закрываются защитными железобетонными пробками и завариваются.
Теплосъем осуществляется посредством циркуляции воды из бассейна, образующего первый контур, через теплообменник. Вода в бассейне нагревается до 9 0 °С, подается насосами в теплообменник, где охлаждается до 70°С , и возвращается в бассейн. Отвод тепла ведется оборотной водой, нагревающейся во втором контуре с 30°С до 80°С .
Водород, выделяющийся за счет радиолиза воды, разбавляется до взрывобезопасных концентраций воздухом, продуваемым над поверхностью воды, и выбрасывается в атмосферу через трубу. Воздух очищается на фильтрах на входе в бассейн и при сбросе из бассейна перед вентилятором. Хотя охлаждающая вода не соприкасается с поверхностью емкостей, предусматривается аварийная очистка воды в количестве 5% от часового оборота воды первого контура, так как нельзя полностью исключить возможность проникновения воды внутрь пеналов при каких-либо нарушениях их герметичности.
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О5 1 03 104 2 3 4 5 qv. Вт/м3
Рис. 6. Зависимость капитальных затрат от удельного тепловыделения остеклованных отходов:
1 — воздушное охлаждение остеклованных отходов при непосредственном контакте охлаждаю
щего воздуха со стенками емкостей, содержащих тепловыделяющий продукт; 2 — воздушное
охлаждение остеклованных отходов через стенки ограждающих труб; 3 — водяное охлаждение
остеклованных отходов через стенки пеналов, содержащих емкости.
3. ТЕХНИКО-ЭКОНОМИЧЕСКАЯ ОЦЕНКА МЕТОДОВ ХРАНЕНИЯОСТЕКЛОВАННЫХ ОТХОДОВ
Оценка методов хранения отходов проводилась для трех вариантов охлаждения:
1) воздушное охлаждение остеклованных отходов при непосредственном контакте охлаждающего воздуха со стенками емкостей, содержащих тепловыделяющий продукт;
2) воздушное охлаждение остеклованных отходов через стенки ограждающихтруб;
3) водяное охлаждение остеклованных отходов через стенки пеналов, содержащих емкости.
Результаты расчета капитальных затрат на строительство хранилищ остеклованных отходов приведены на рис.6. Наименьшие капитальные затраты на сооружение хранилища получены для третьего варианта благодаря большой компактности хранилища. Наибольшие капитальные затраты будут иметь место при сооружении хранилища по первому варианту для продуктов с удельным тепловыделением, равным 4 - 104 Вт/м3, так как большая, чем во втором варианте, компактность хранилища не компенсирует затрат на сооружение фильтровальной станции.
Во всех случаях увеличение концентрирования отходов при остекловывании позволяет снижать затраты, однако во втором варианте при удельном тепловыделении, равном 2-104- 105 Вт/м3, наблюдается небольшой рост расходов вследствие большего удельного расхода конструкционного материала на единицу захораниваемого тепловыделяющего продукта.
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5-103 104 2 3 4 5 qv, Вт/м3
Рис. 7. Зависимость приведенных затрат от удельного тепловыделения остеклованных отходов:
1 — воздушное охлаждение остеклованных отходов при непосредственном контакте охлаждающего
воздуха со стенками емкостей, содержащих тепловыделяющий продукт; 2 — воздушное охлаж
дение отходов через стенки ограждающих труб; 3 — водяное охлаждение остеклованных отходов
через стенки пеналов, содержащих емкости.
Годовые эксплуатационные расходы на содержание хранилища меньше всего во втором варианте, при этом основную долю затрат составляют амортизационные отчисления, затраты на емкости и энергетику. В первом варианте эксплуатационные затраты увеличиваются в основном за счет необходимости замены фильтров. Наибольшие эксплуатационные расходы были установлены для хранилища с водяным охлаждением. В этом случае самый большой вклад вносят расходы на очистку воды, затраты на пеналы и амортизационные отчисления.
В соответствии с принятой в СССР методикой [1], было сделано сравнение рассматриваемых методов хранения по капитальным затратам. Полученные результаты представлены на рис.7. Видно, что по приведенным затратам оба варианта воздушного охлаждения мало отличаются друг от друга, т.е. экономическая оценка не может быть определяющим фактором при выборе того или иного способа отвода тепла. Такой выбор должен делаться из технологических соображений. Приведенные затраты для варианта с водяным охлаждением в рассмотренном диапазоне удельного тепловыделения остеклованных отходов несколько выше, чем при воздушном охлаждении,
150 КОНДРАТЬЕВ и др.
но по мере роста концентрирования жидких отходов при остекловывании происходит
заметное снижение приведенных затрат. Это говорит о целесообразности использова
ния такого метода охлаждения при удельном тепловыделении выше 105 Вт/м3.
Рассмотренная экономическая оценка относится к условиям хранения остекло
ванных продуктов с меньшей термической устойчивостью. При хранении продуктов
с более высокой допустимой температурой хранения размеры емкостей или удельное
тепловыделение могут быть увеличены, благодаря чему экономические показатели
хранения таких отходов улучшаются.
4. ВЫВОДЫ
1. Проведенные расчеты допустимых емкостей доя остеклованных отходов по
казали, что при наличии отходов с удельным тепловыделением 5 • 103- 105 Вт/м3 могут
применяться все рассмотренные конструкции хранилищ с воздушным и водяным ох
лаждением. Однако, в случае применения воздушного охлаждения тепловыделяющих
продуктов через ограждающие трубы допустимые размеры емкостей приходится силь
но ограничивать при хранении отходов с низкой термической устойчивостью и удель
ным тепловыделением, превышающим 5 • 104 Вт/м3.
2. Экономическся оценка приведенных капитальных затрат показала, что в рас
смотренном диапазоне удельного тепловыделения отходов показатели для обоих ва
риантов воздушного охлаждения мало отличаются друг от друга, и выбор схемы ох
лаждения может производиться исходя из технологических условий. Увеличение сте
пени концентрирования отходов в процессе остекловывания и увеличение термичес
кой устойчивости остеклованных отходов позволяют снизить приведенную стоимость
хранения отходов. Очевидно, что переход от стеклоподобных к базальтоподобным
материалам повысит надежность и снизит стоимость их хранения.
3. Небольшой рост затрат при удельном тепловыделении 2 • 104 - 10s Вт/м3 для
хранилищ с воздушным охлаждением через стенки ограждающих труб связан с конст
руктивными особенностями выбранного хранилища.
4. Приведенные затраты для хранилищ с водяным охлаждением в рассмотрен
ном диапазоне удельного тепловыделения несколько выше, чем для хранилищ с воз
душным охлаждением, однако при увеличении степени концентрирования отходов в
процессе остекловывания наблюдается быстрое снижение расходов. Хранилища с
водяным охлаждением целесообразнее применять при удельном тепловыделении
остеклованных отходов, превышающем 105 Вт/м3.
ЛИТЕРАТУРА
[ 1] Типовая Методика Определения Экономической Эффективности Капитальных Вложений, Экономика, М., 1969.
UNDERGROUND DISPOSAL OF LIQUID WASTE,
DISPOSAL OF SOLID WASTE AT SHALLOW DEPTH
AND IN ROCK CAVERNS
(Session III)
N.S. SUNDER RAJAN
India
Chairman
IAEA-SM-243/110
ПРИНЦИПЫ
ОЦЕНКИ НАДЕЖНОСТИ ПОДЗЕМНОГО ЗАХОРОНЕНИЯ
РАДИОАКТИВНЫХ ЖИДКИХ ОТХОДОВ
В ГЛУБОКИЕ ГЕОЛОГИЧЕСКИЕ ФОРМАЦИИ
И ПУТИ ЕЕ ПОВЫШЕНИЯ
О.Л. КЕДРОВСКИЙ, М.К. ПИМЕНОВ, Н.А. РАКОВ,
А.И. РЫБАЛЬЧЕНКО, Ф.П. ЮДИН
Государственный комитет
по использованию атомной энергии СССР,
Москва,
Союз Советских Социалистических Республик
Abstract-Аннотация
PRINCIPLES FOR EVALUATING THE RELIABILITY OF UNDERGROUND DISPOSAL
OF LIQUID RADIOACTIVE WASTES IN DEEP GEOLOGICAL FORMATIONS AND
WAYS OF IMPROVING RELIABILITY.
Because of the rapid development of nuclear power, protection of the environment
against contamination by radioactive wastes has become a task of major importance. As a
result, studies have been performed and new techniques have been developed for the manage
ment and disposal of radioactive wastes. One of these techniques involves disposal in
geological formations by pumping liquid radioactive wastes down boreholes into deep-lying
permeable confining beds, which can be relied on to remain isolated from the earth’s surface
and from the groundwater which passes into the water supply. An underground disposal
system can be set up if the isolation of wastes can be proved to be reliable and safe for both
present and future generations. One of the main considerations involved in determining the
reliability of the method of disposal is the estimated isolating capacity of the geological
medium for use as a repository for radioactive wastes. Engineering techniques can now ensure
to a sufficient degree the reliability of the process equipment of underground repositories
and the integrity of the boreholes. Evaluation and demonstration of the reliability and safety
of a disposal method involve analysis of the possible pathways of radionuclides into the
biosphere, calculation of the migration parameters of nuclides and evaluation of the role of
barriers to migration. For comparing the relative reliability of different waste disposal methods,
quantitative criteria of reliability should be used, the most important of which is the
probability or risk of radioactivity leaking from the disposal system into the environment.
When estimating the level of risk, structured statistical models of the sector of the geological
medium containing the repository, the results of a statistical analysis of geological
features and nuclide migration parameter forecasts are used. The estimates obtained can be
used for specifying the requirements with respect to geological structure and to the
characteristics of the sector which is to contain the repository, when a comparison is made
between different sectors. In order to increase the dependability and safety of underground
disposal, a reliability programme embracing all stages of the repository’s construction and
operation should be drawn up and strictly applied. This programme would include the
153
154 КЕДРОВСКИЙ и др.
optimization of underground repository operation with automated surveillance and control
systems. If the various tasks which need to be carried out to ensure the integrity of the
underground repository are correctly performed when it is constructed, a high degree of
safety in radioactive waste disposal will be achieved.
ПРИНЦИПЫ ОЦЕНКИ НАДЕЖНОСТИ ПОДЗЕМНОГО ЗАХОРОНЕНИЯ РАДИОАКТИВНЫХ ЖИДКИХ ОТХОДОВ В ГЛУБОКИЕ ГЕОЛОГИЧЕСКИЕ ФОРМАЦИИ И ПУТИ ЕЕ ПОВЫШЕНИЯ.
Интенсивное развитие ядерной энергетики выдвинуло на первый план проблему охраны окружающей среды от загрязнения радиоактивными отходами и обусловило проведение исследований и разработок новых способов обезвреживания и удаления радиоактивных отходов. Одним из таких способов является захоронение радиоактивных отходов в геологические формации путем нагнетания жидких отходов через буровые скважины в глубокозалегающие проницаемые пласты-коллекторы, надежно изолированные от земной поверхности и подземных вод, используемых в хозяйственной деятельности. Создание систем подземного захоронения осуществляется в том случае, если доказана надежность изоляции отходов и их безопасность как для ныне существующих, так и для будущих поколений. Центральным вопросом надежности захоронения является оценка изолирующей способности геологической среды как вместилища радиоактивных отходов. Надежность технологического оборудования подземных хранилищ и скважин может быть обеспечена на достаточно высоком уровне инженерными методами. В процессе оценки и обоснования надежности и безопасности захоронения анализируются вероятные пути поступления радиоактивных нуклидов в биосферу, рассчитываются параметры миграции нуклидов и оценивается роль барьеров на пути миграции. Для сравнения надежности различных способов удаления отходов целесообразно применять количественные показатели надежности, основным из которых является вероятность выхода радиоактивности из системы захоронения в окружающую среду или уровень риска. При оценке уровня риска используются структурные статистические модели участка геологической среды, в которой размещено хранилище, результаты статистического анализа геологических признаков и прогнозных параметров миграции нуклидов. Полученные оценки могут быть использованы при формулировании требований к геологическому строению и характеристикам участков размещения хранилищ, при сопоставлении различных участков их предполагаемого размещения. С целью повышения надежности и безопасности подземного захоронения целесообразно предусматривать составление и строгое выполнение программы надежности, охватывающей все стадии создания и эксплуатации хранилища. В состав программы входит оптимизация эксплуатации подземных хранилищ, осуществляемая на основе автоматизированных систем контроля и управления. Выполняемый при создании подземных хранилищ комплекс работ по обеспечению надежности захоронения позволяет достигнуть высокой безопасности захоронения радиоактивных отходов.
1. ОБЩИЕ ПОЛОЖЕНИЯ
Захоронение отходов в геологические формации является одним из наиболее перс
пективных способов локализации и обезвреживания радиоактивных отходов крупных
исследовательских центров, ядерных энергетических установок и установок по перера
ботке облученного топлива. Захоронение радиоактивных отходов в геологические фор
мации наиболее полно отвечает рекомендациям Международной комиссии по радиоло
гической защите (МКРЗ) об обеспечении минимально возможной дозы облучения насе
ления, обусловленной радиоактивными отходами, так как позволяет изолировать отхо
ды в геологической формации от среды обитания человека на период времени, необхо
димый для снижения уровня активности отходов до безопасных пределов.
IAEA-SM-243/110 155
В различных странах в зависимости от природных условий и географического по
ложения, традиций, плотности населения и уровня развития соответствующих отраслей
науки и техники исследуются, разрабатываются и находят применение различные спосо
бы захоронения.
В СССР признаны перспективными следующие направления в развитии способов
подземного захоронения радиоактивных отходов:
— захоронение отвержденных и твердых отходов в горные выработки;
— захоронение жидких отходов низкого и среднего уровней активности в глубо-
козалегающие пористые пласты-коллекторы.
Захоронение жидких отходов осуществляется путем их нагнетания в пласт-коллек
тор через систему нагнетательных скважин; для нагнетания выбираются пласты и водо
насыщенные комплексы, расположенные в зоне застойного режима и надежно изолиро
ванные от вод, используемых в хозяйственной деятельности. Захоронение является
контролируемым и управляемым процессом.
В СССР имеется положительный опыт захоронения в глубоких поглощающих го
ризонтах жидких отходов низкого и среднего уровней активности [1,2].
Обязательным требованием, предъявляемым к системам подземного захороне
ния радиоактивных отходов, является надежность и безопасность захоронения. Под
надежностью подземного захоронения понимается свойство системы захоронения обес
печивать удаление отходов в подземное хранилище и их изоляцию от биосферы в тече
ние необходимого периода времени, определяемого периодом распада до безопасного
уровня наиболее опасного и долгоживущего радиоактивного нуклида, содержащегося
в отходах. Надежность захоронения отходов является решающим фактором, от которо
го зависит возможность строительства подземных хранилищ в геологических формациях.
Создание систем подземного захоронения в геологических формациях осуществляется
в том случае, если доказана надежность изоляции отходов и их безопасность как для
ныне существующих, так и для будущих поколений людей.
Большое значение, придаваемое надежности захоронения, обусловлено тем, что
радиоактивные отходы, в какой бы форме они не были (жидкие, твердые или газообраз
ные) и где бы они не находились (в емкостях на поверхности, в глубоких пластах-кол
лекторах, в соляных шахтах или других горных выработках), представляют потенциаль
ную опасность. В каждом случае могут быть смоделированы более или менее вероят
ные ситуации, при которых возможно поступление отходов в биосферу.
Оценка надежности подземного захоронения может быть как качественной, так и
количественной и включает в себя:
— анализ возможных путей миграции радиоактивных нуклидов — компонентов
отходов из подземного хранилища, расположенного в геологических формациях,
в биосферу;
— расчет параметров возможной миграции нуклидов ;
— оценку надежности поверхностного оборудования подземных хранилищ;
— оценку надежности буровых скважин;
— определение количественных показателей надежности захоронения.
156 КЕДРОВСКИЙ н др.
ПОВЕРХНОСТНОЕ ТЕХНОЛОГИЧЕСКОЕ ОБОРУДОВАНИЕ ДЛЯ ПРИЕМА,
ПОДГОТОВКИ И ТРАНСПОРТИРОВКИ ОТХОДОВ
БУРОВЫЕ СКВАЖИНЫ РАЗЛИЧНОГО НАЗНАЧЕНИЯ
ПЛАСТ-КОЛЛЕКТОР,БУФЕРНЫЙ ГОРИЗОНТ,
ИЗОЛИРУЮЩИЕ ВОДОУПОРНЫЕ ГОРИЗОНТЫ
Рис. 1. Функциональная схема подземного хранилища ж идких радиоактивных отходов.
Количественные показатели надежности являются формализованными критери
ями, содержащими информацию о надежности в наиболее краткой форме. Важным
показателем надежности системы захоронения является также готовность системы,
оцениваемая отношением периода времени, в течение которого система выполняет
свои функции, к общему времени эксплуатации системы.
Показатели надежности используются при принятии решения об осуществлении
захоронения, о выборе наиболее безопасной схемы хранилища.
Проведение детальной оценки надежности захоронения вызывается необходи
мостью создания оптимальных систем захоронения, сочетающих техническую осуществи
мость, экономическую рентабельность и разумный уровень риска, а также требовани
ем строгого доказательства безопасности захоронения перед директивными и контро
лирующими государственными органами.
Надежность захоронения в геологические формации необходимо рассматривать
для ограниченного периода времени (а не для бесконечности, что бессмысленно, так
как изменению в конечном итоге подвергнутся все геологические формации). Надеж
ность рассматривается для периода, в течение которого отходы представляют опреде
ленную опасность. Этот период определяется временем радиоактивного распада нук
лидов до безопасного уровня, который находится на основе различных предпосылок.
Например, безопасный уровень может соответствовать концентрациям нуклидов, уста
новленным санитарными органами как допустимым в водах открытых водоемов или
таким концентрациям в подземном хранилище, что вероятный выход нуклидов из хра
нилища возможен в концентрациях, не выше допустимых.
IAEA-SM-243/110 157
Элементы функциональной схемы подземного хранилища жидких радиоактив
ных отходов в геологических формациях могут быть разбиты на 3 основных груп
пы (рис. 1).
При оценке безопасности захоронения анализ надежности выполняется для каж
дой из этих групп.
Технологические системы на поверхности и скважины являются инженерными со
оружениями, надежность которых может быть обеспечена на достаточно высоком уров
не применением соответствующего оборудования, конструкций и материалов. Высокая
степень ремонтопригодности и резервирования оборудования и скважин, предусматрива
емая при проектировании, повышает надежность их эксплуатации. Участок недр, выбран
ный для создания хранилища, является природным объектом. Геологическое строение
недр и основные характеристики горизонтов в области влияния подземного хранилища
не могут быть откорректированы инженерными методами, имеются ограничения в об
следовании недр. Поэтому при обсуждении проблем безопасности захоронения прежде
всего необходимо рассмотреть надежность изоляции радиоактивных отходов от био
сферы за счет свойств геологической среды.
Этот вопрос является одним из принципиальных и дискутируемых в настоящее
время вопросов проблемы подземного захоронения.
2. АНАЛИЗ ВОЗМОЖНЫХ ПУТЕЙ МИГРАЦИИ НУКЛИДОВ ИЗ ХРАНИЛИЩА И
РОЛЬ ПРИРОДНЫХ БАРЬЕРОВ
Подземное хранилище жидких радиоактивных отходов представляет собой учас
ток недр, содержащий пористый или пористо-трещинный водонасыщенный пласт-кол
лектор, в который осуществляется нагнетание отходов, буферный горизонт, подстила
ющие и перекрывающие пласт-коллектор водоупоры.
В качестве пласта-коллектора целесообразно выбирать горизонты, содержащие
древние высокоминерализованные воды,не используемые в хозяйственной деятельности
и практически полностью изолированные от дневной поверхности и водоносных гори
зонтов, используемых человеком. Такие горизонты встречаются, как правило, в оса
дочных чехлах платформенных областей и в краевых прогибах. Пласты-коллекторы
выбираются таким образом, чтобы выше них имелся буферный горизонт — пористый
пласт, отделенный от пласта-коллектора водоупором.
Отходы, нагнетаемые через одну или несколько скважин, поступают в пласт-кол-
лектор, вытесняют пластовую жидкость из порового пространства и занимают опреде
ленный объем. На границе отходы—пластовая жидкость возникает переходная зона
(зона дисперсии). Размер и форма объема пород, занятых отходами в пласте-коллек
торе, зависят от расхода нагнетания, эффективной пористости пород и мощности плас
та-коллектора, фильтрационной неоднородности. Под влиянием естественного потока
подземных вод происходит смещение контура отходов. При фильтрации отходов в
пласте-коллекторе происходят физико-химические взаимодействия компонентов от
ходов с породами, в результате которых значительная часть долгоживущих радиоак
158 КЕДРОВСКИЙ и др.
тивных нуклидов задерживается породами пласта-коллектора и мигрирует со значи
тельно меньшими скоростями, чем невзаимодействующие компоненты.
Выход радиоактивных нуклидов из подземного хранилища возможен при дости
жении отходами, мигрирующими в водоносном горизонте, областей разгрузки или пе
ретекания в вышележащие водоносные горизонты, при вертикальной фильтрации от
ходов и диффузионном массопереносе через перекрывающие пласт-коллектор водоу-
поры, в результате сейсмической активности или случайного вскрытия пласта-коллек
тора при буровых или горных работах.
Прогнозные расчеты миграции отходов и входящих в их состав радиоактивных
нуклидов по наиболее опасным направлениям и по вертикали через водоупоры выпол
няются для различных периодов времени, в том числе и для периода, необходимого
для распада нуклидов до безопасного уровня. Расчеты основываются на теоретичес
ких решениях задач гидродинамики и массопереноса в геологической среде и прово
дятся с использованием ЭВМ (аналоговое моделирование). Исходными данными для
расчетов являются характеристики геологического разреза и физические свойства гор
ных пород, химические составы отходов и подземных вод, направление и скорость ес
тественного потока. При расчетах учитывается разность плотностей отходов и подзем
ных вод, вертикальная и плановая фильтрационная неоднородность, взаимодействие
отходов с породами пласта-коллектора.
Все эти данные должны быть получены по результатам геолого-разведочных ра
бот, лабораторных и полевых исследований, которые предваряют создание хранилища.
До начала расчетов необходимо установить геолого-гидрогеологическую схему райо
на хранилища, иметь данные о региональных гидрогеологических условиях.
2.1. М играция отходов по пласту-коллектору
Максимальное распространение отходов в пласте-коллекторе в направлении естест-
Li L2венного потока рассчитывается с использованием выражения: R = г + — + — , где
R - расстояние контура от нагнетательной скважины; г — максимальное распростране
ние отходов от нагнетательной скважины в направлении естественного потока без уче
та взаимодействия отходов с породами и разности плотностей (поршневое вытеснение) ;
L t- ширина зоны гидравлической дисперсии; L 2 - ширина переходной зоны между
отходами и пластовыми водами, если плотность их различна.
Для бесконечного по площади однородного пласта величина г может быть рассчи
тана по следующей формуле [3] :
О - 1, , к-у- + r0 + J — O í + t í ) (1)J m n 0 n0 v '
где Q —расход нагнетания отходов; t i —продолжительность нагнетания отходов;
h —продолжительность миграции отходов после окончания эксплуатации хранилища;
m —эффективная мощность пласта-коллектора; п0 —эффективная пористость пласта-кол
лектора; г0 — приведенный радиус системы нагнетательных скважин; J — уклон естест
венного потока подземных вод; к — коэффициент фильтрации.
IAEA-SM-243/110 159
Ширина зоны дисперсии Ь1; между точками зоны с концентрациями нуклида, сос
тавляющими 0,01 и 0,99 концентрации в отходах, определяется по выражению [4] :
L i=4,7 а / — ( t , + t j ) (2)V n0
где D - коэффициент гидравлической дисперсии.
В работе [4] приведены также расчетные выражения и номограммы для определе
ния ширины зоны дисперсии между точками с различными относительными концентра
циями компонентов отходов, распространяющихся в подземных водах.
Если при расчетах миграции необходимо учесть накопление нуклида породами,
вместо пористости п0 в вышеприведенные выражения необходимо подставить эффек
тивную пористость пэ с учетом накопления нуклида
пэ = по + р (3)
где Р — коэффициент распределения, равный отношению количества нуклида, накопив
шегося в единице объема породы, к объемному содержанию нуклида в равновесном
растворе. Величина L 2 для точек со значениями избыточной плотности 0,01 и 0,99 рассчи
тывается по выражению [4] :
L j=3 ,97 л/к-т-Д?— (t, + t2) (4)V n0
У “ Угде А у = —-- — (у и у . — п л о т н о с т и подземных вод и отходов, соответственно).
? 2
Имеются подобные выражения и для расчета положения контура в направлениях
от нагнетательной скважины, не совпадающих с естественным потоком.
В результате расчетов получают сведения о местоположении контура отходов для
различных периодов времени. Достаточное удаление контура отходов от областей раз
грузки или перетекания для периода времени, необходимого для распада нуклидов до
установленного уровня, дает возможность оценить надежность захоронения,как высокую.
2 .2 . Вертикальная миграция отходов
Разность напоров между уровнями подземных вод в естественном режиме двух
или более водоносных горизонтов, залегающих друг над другом, свидетельствует об
изоляции горизонтов или их затрудненном водообмене. При нагнетании отходов про
исходит нарушение естественного гидродинамического режима, которое может явиться
причиной возникновения вертикальной фильтрации через водоупор, прекращающейся
после окончания нагнетания и восстановления естественного режима горизонтов.
Время от начала негнетания, после которого возможно поступление отходов в
вышележащий горизонт, определяется по формуле [5] :
_ n'-k-m-(m')2
t0 2 0 ? (5)
160 КЕДРОВСКИЙ и др.
где Q — расход нагнетания отходов; к и т — коэффициент фильтрации и мощность пласта-коллектора, соответственно; k’, m’, n' — коэффициент фильтрации, мощность и пористость перекрывающего водоупора, соответственно.
Объем отходов, поступающих в вышележащий горизонт через время t0, определяется по формуле [5] :
k'-R2Qn ; т г ( Р - Р ' ) - ( l nR- in^-aS)
2k m (6)
где R — радиус контура отходов на данный момент времени t работы нагнетательных скважин; г0 — приведенный радиус нагнетательных скважин; Р — приведенный напор в нагнетательных скважинах на момент t; F — приведенный напор в вышележащем горизонте.
Очевидно, для исключения поступления отходов в вышележащие горизонты необходимо, чтобы значения расхода нагнетания Q, время эксплуатации хранилища и характеристики разреза удовлетворяли определенным требованиям, которые могут быть сформулированы на основании прогнозных расчетов.
В естественном гидродинамическом режиме, нарушенном нагнетанием отходов, расход фильтрации через 1м2 водоупора может быть приближенно оценен по формуле [5] :
ДРg ~ k —— (7)
m
где ДР — разность напоров между платом-коллектором и вышележащим горизонтом.Анализ гидродинамических и гидрогеохимических условий, результатов определения
возраста подземных вод показывают, что глинистые водоупоры характеризуются хорошими изолирующими свойствами и поступление нуклидов через водоупор является маловероятным [6]. Высокие сорбционные свойства глин также способствуют изоляции нуклидов от вышележащих водоносных горизонтов.
Буферный горизонт, залегающий выше пласта-коллектора и отделенный от него водоупором, является резервным горизонтом. Поступление нуклидов в буферный горизонт возможно только в малых концентрациях ввиду вышеперечисленных факторов.В буферном горизонте произойдет снижение концентраций в результате разбавления, поэтому проникновение нуклидов выше буферного горизонта также маловероятно. При обнаружении в буферном горизонте отходов в процессе контроля захоронения могут быть своевременно проведены необходимые мероприятия по прекращению вертикальной фильтрации.
Важной характеристикой подземных вод, позволяющей прогнозировать масштабы миграции отходов, является относительный или абсолютный возраст подземных вод, определяемый при проведении гидрохимических изотопных исследований и гелиевой съемки. Значения абсолютного возраста используются для определения скорости естественного потока, различия возраста вод нескольких горизонтов одного комплекса свидетельствуют об их разобщенности [7].
Диффузионный массоперенос нуклидов рассчитывается путем решения уравнений массопереноса с использованием коэффициентов диффузии, определенных в лаборато-
IAEA-SM-243/110 161
рии. Диффузионный массоперенос характеризуется низкими скоростями миграции (например, время диффузии стронция-90 через глинистый водоупор мощностью 10 м составляет ~ 1000 лет) .
Расчеты вертикальной фильтрации и массопереноса позволяют оценить возможность поступления отходов в вышележащие горизонты и сформулировать в каждом случае требования к характеристикам геологического разреза, удовлетворяющим условиям надежного захоронения.
Подводя итог сказанному относительно возможности вертикальной миграции отходов, можно предположить, что такая миграция для сплошных глинистых водоупоров будет незначительна и не повлияет на надежность захоронения. Однако это не так в случае наличия ’’дефектов” в водоупорах, например, литологических окон, в которых глинистые породы замещаются проницаемыми, фильтрующих зон тектонических нарушений, трещиноватых зон. Такие ’’дефекты” могут быть обнаружены известными методами при геологоразведочных работах, при выявлении ’’дефектов” подземное хранилище на данном участке, как правило, не создается.
2.3. Радиогенный разогрев, влияние геологических процессов и хозяйственной
деятельности.
При захоронении отходов высокого уровня активности отмечается разогрев пласта-коллектора, заполненного отходами, в результате радиогенного тепловыделения. -В случае разогрева пласта-коллектора до температуры, превышающей температуру парообразования в пластовых условиях, увеличивается потенциальная опасность эксплуатации хранилища. В связи с этим в оценку безопасности и надежности захоронения отходов высокого уровня активности входят расчеты температуры разогрева, выполняемые путем численных решений теплофизических уравнений с использованием ЭВМ. Получаемые результаты позволяют сформулировать требования к режиму нагнетания и величине максимальной удельной активности отходов, при которых разогрев пласта не превысит установленных значений.
Геологическими процессами, в результате которых возможен выход радиоактивных нуклидов из хранилища, являются сейсмические явления, нарушающие сплошность водоупоров. Для исключения этих явлений подземные хранилища располагаются в сейсмически неактивной зоне. По предварительной оценке подземное захоронение возможно в районах сейсмической активности не более 7 баллов по шкале Рихтера.
Для исключения случайного вскрытия пласта-коллектора, заполненного отходами, при бурении скважин или проведении горных работ, а также для предотвращения изменений режима водоносного комплекса, которые могут отрицательно повлиять на безопасность эксплуатации хранилища, в районе размещения хранилища предусматривается зона санитарной охраны, цель которой — контроль за проведением всех мероприятий в районе хранилища, введение ограничений на использование недр в различных целях.
162 КЕДРОВСКИЙ н др.
2.4. Защитные барьеры для радиоактивных отходов
Под барьерами для радиоактивных отходов, находящихся в подземном хранилище, понимаются такие геологические образования, инженерные сооружения или процессы, которые исключают выход нуклидов из хранилища. Для жидких радиоактивных отходов в подземном хранилище такими барьерами являются:
— водоупорные породы, перекрывающие поглощающий пласт-коллектор;— буферный горизонт;— низкие скорости потока подземных вод и застойный режим водонасыщенного
комплекса, в котором выбран пласт-коллектор, что обеспечивает распад нуклидов за время их пребывания в недрах;
— задержка миграции нуклидов в недрах в результате их взаимодействия с породами пласта-коллектора;
— специальные конструкции скважин и применяемые материалы для их сооружения, предохраняющие вышележащие водоносные горизонты от загрязнения;
— контроль и управление подземным захоронением, комплекс противоаварийных мероприятий.
Защитная роль каждого барьера оценивается с использованием вышеприведенных расчетных выражений, на основе анализа эффективности управления захоронением и противоаварийных мероприятий.
Действенность защитных функций барьеров й надежность подземного захоронения в целом может быть оценена количественными показателями надежности.
3. ПРИНЦИПЫ КОЛИЧЕСТВЕННОЙ ОЦЕНКИ НАДЕЖНОСТИ ЗАХОРОНЕНИЯ
РАДИОАКТИВНЫХ ЖИДКИХ ОТХОДОВ
Количественная оценка надежности захоронения выполняется на основе анализа статистических моделей геологической среды и протекающих в хранилище и его окружении процессов.
Целесообразность вероятностного подхода к оценке надежности захоронения вытекает из дисперсии физических свойств и характеристик поглощающих пластов-коллекторов и водоупоров, статистического характера поля скоростей потока отходов и подземных вод в горизонте, а также из того, что пласт-коллектор и перекрывающие водоу- поры являются природными объектами, относительно малодоступными для непосредственной ревизии, свойства которых известны с некоторой степенью неопределенности.
Одним из возможных путей количественной оценки надежности захоронения является расчет вероятности выхода радиоактивных отходов из хранилища в окружающую среду [8].
Выход радиоактивных отходов из хранилища может произойти в том случае, если при расчетах хранилища и процессов миграции были допущены неточности, обусловленные отличием значений физических свойств пород и характеристик горизонтов, принятых при расчетах, от фактических значений.
IAEA-SM-243/110 163
Физические свойства пород и характеристики горизонтов, определяемые при геологоразведочных работах, лабораторных исследованиях и по фондовым материалам, представляют собой выборку из генеральных совокупностей всех возможных значений свойств и характеристик. Рассеяние этих значений определяется как естественной геологической изменчивостью, так и случайной ошибкой определения.
При прогнозных расчетах принимаются как наиболее вероятные средние значения свойств и характеристик, их рассеяние оценивается дисперсией. Эти значения не являются истинными, а лишь оценочными, так как получены по ограниченному числу наблюдений. Более правильно эти значения характеризовать доверительным интервалом, в пределах которого находятся истинные значения. Пределы интервала зависят от принятого уровня значимости (доверительной вероятности) , числа наблюдений, по которым определяется среднее значение.
Очевидно, что теоретически могут существовать средние значения характеристик, при которых надежность захоронения не обеспечивается и возможен выход радиоактивных нуклидов в окружающую среду. Вероятность существования таких значений и, соответственно, вероятность выхода нуклидов принимается как ’’ненадежность” или ’’уро
вень риска” захоронения — Н. Надежность захоронения М = 1 - Н.Для определения значений М или Н имеющийся массив определений физических
свойств или характеристик аппроксимируется нормальным или логнормальным распределением, доверительные интервалы и вероятности определяются с использованием t-распределения. Предельные значения свойств и параметров, при которых надежность захоронения не будет обеспечена и вероятность существования которых необходимо определить, находятся с использованием выражений, приведенных в предыдущем разделе, для конкретных геологических условий.
Физическими свойствами и характеристиками горизонтов,используемыми при прогнозных расчетах, являются: коэффициент фильтрации, пористость, теплофизические свойства и т.д. Расчет надежности необходимо проводить по каждому из невзаимосвязанных свойств или характеристик. Суммарный уровень риска оценивается выражением:
H£ = ¿ H i - ¿ (Hi- Hj)i=l i, j=l, уф)
где H! , . . Hj , . . Hn — уровень риска по каждому свойству; n — количество свойстви характеристик, 1< j, i < п.
В том случае, если наблюдается закономерное изменение свойств или характеристик в пространстве хранилища (тренд), для определения средних значений пространство разделяется на участки, в пределах которых тренд отмечается слабо. Расчет надежности ведется для каждого из участков. Могут быть также применены специальные приемы учета влияния тренда на основе регрессионного и дисперсионного анализов [9]. Если в хранилище имеется объект с аномальными свойствами, например в пласте-коллекторе — слой максимальной проницаемости среди слоев низкой проницаемости, прогноз миграции и расчет надежности выполняется для каждой группы слоев, выделенных по проницаемости.
16 4 КЕДРОВСКИЙ и др.
При расчетах надежности по некоторым характеристикам целесообразно вместо распределений средних значений характеристик рассматривать непосредственно распределение серии определений или измерений.
Расчет надежности ведется для периода времени, в течение которого отходы представляют опасность для биосферы.
Полученные в результате расчетов значения надежности или уровни риска сопоставляются со значениями, принятыми органами, регулирующими или контролирующими системы обезвреживания отходов, ядерных установок и крупных инженерных сооружений, а также используются при сравнении надежности нескольких установок или систем, по обезвреживанию или захоронению отходов. В том случае, если регулирующими органами установлен определенный уровень надежности для подземного хранилища, предлагаемая концепция оценки надежности используется для формулирования требований к параметрам геологического разреза и объему геологоразведочных и лабораторных работ для исследования условий захоронения.
Проведенные расчеты для существующих подземных хранилищ жидких радиоактивных отходов показывают высокую надежность захоронения. Уровень риска или вероятность выхода радиоактивности в окружающую среду из-за дисперсии физических свойств пород или характеристик разреза оценивается незначительными величинами.
Применение вероятностного подхода к оценке надежности захоронения возможно в том случае, если имеется достаточное количество данных для статистической обработки. На начальной стадии работ по созданию хранилища имеется ограниченное число таких данных, которые иногда принимаются только по фондовым материалам. При этом целесообразно применить ’’метод худшего случая”, заключающийся в проведении расчетов с использованием наиболее неблагоприятных значений параметров [9]. Полученные результаты позволяют сделать вывод о целесообразности постановки дальнейших исследований в данном районе или уточнить объем необходимых разведочных работ.
При анализе потенциальных ошибок прогнозных расчетов миграции нуклидов и оценке надежности не следует исключать возможность ошибочных моделей геолого-гид- рогеологических условий и процессов в хранилище. Например, если при построении модели не учитывались фильтрационные окна в водоупорах, зоны аномально-высокой проводимости и т.д. Такие ошибки обесценивают прогнозы миграции и расчеты надежности. Ошибки могут быть полностью исключены при правильной постановке и достаточном объеме исследовательских работ.
4. ОЦЕНКА НАДЕЖНОСТИ ОБОРУДОВАНИЯ НА ПОВЕРХНОСТИ И БУРОВЫХ
СКВАЖИН
Поверхностное оборудование характеризуется значительной простотой по сравнению, например, с оборудованием по остекловыванию отходов. При проектировании,
IAEA-SM-243/110 165
изготовлении и монтаже к качеству оборудования предъявляются высокие требования, не меньшие, чем к ядерным установкам, в связи с чем надежность оборудования может быть обеспечена весьма высокой. Оценка надежности может быть выполнена известными приемами (’’дерево отказов”) , применяемыми при оценке надежности ядерных энергетических установок,и в данном докладе не рассматривается [8, 10].
Основными типами буровых скважин подземных хранилищ радиоактивных жидких отходов являются: нагнетательные, через которые осуществляется нагнетание отходов, и контрольно-наблюдательные, используемые для контроля захоронения.
Вероятными ’’отказами” скважин, которые могут привести к загрязнению окружающей среды, являются: нарушение целостности обсадной колонны в результате ее коррозии и разрушение цементного кольца в затрубном пространстве обсадной колонны.
Оценку надежности функционирования скважин можно осуществлять путем статистического анализа отказов и вызвавших их причин. В практике эксплуатации подземных хранилищ радиоактивных жидких отходов в СССР практически отсутствуют подобные случаи отказов скважин.
Возможным является статистический анализ отказов нефтяных и водозаборных скважин. Однако их конструкции, применяемые материалы и технология проводки значительно отличаются от скважин, применяемых на подземных хранилищах радиоактивных отходов, поэтому этот анализ не будет представительным для наших целей. Результаты такого анализа могут быть использованы при создании наиболее надежных конструкций скважин.
Относительно высокая надежность скважин подземных хранилищ радиоактивных отходов обеспечивается применением высококачественных материалов для их сооружения, например, антикоррозионных сталей и сплавов, специальными конструкциями скважин, исключающими разрушение скважин и позволяющими осуществлять контроль за их состоянием, выполнением программы контроля качества всех работ при сооружении скважин, возможностью эффективного и своевременного ремонта, а при необходимости и их ликвидации.
Надежность обеспечивается резервированием скважин, применяемый резерв составляет не менее 50%. Одним из отказов скважин, влияющим на готовность системы захоронения, является кольматация прифильтровой зоны пласта, вызывающая снижение расхода и повышение давления нагнетания отходов. Резервирование скважин и имеющиеся методы борьбы с кольматацией скважин (гидроразрыв, кислотные обработки) позволяют достичь практически стопроцентной готовности системы захоронения.
5. ПУТИ ПОВЫШЕНИЯ НАДЕЖНОСТИ ПОДЗЕМНОГО ЗАХОРОНЕНИЯ
Необходимая надежность подземного захоронения может быть обеспечена путем принятия соответствующих мер на всех стадиях работ по осуществлению захоронения.С этой целью предусматривается программа надежности, содержащая перечень мероприятий по обеспечению надежности, которые необходимо выполнить при проведении гео
166 КЕДРОВСКИЙ и др.
логоразведочных работ и лабораторных исследований, проектировании, строительстве и эксплуатации хранилища, рекомендации по методам исследований, расчетов и контрольных наблюдений.
Эта программа требует уделять особое внимание всем тем вопросам, которые определяют надежность и безопасность захоронения.
При проведении геологоразведочных работ и лабораторных физико-химических исследований такими вопросами, например, являются определение непрерывности распространения перекрывающих пласт-коллектор водоупорных образований по площади и мощности, возраст, скорость и направление потока подземных вод, миграционная способность радиоактивных нуклидов и т.д. Для объективной оценки надежности необходимо составить первоначально качественную, а затем и количественную модель геоло- го-гидрогеологической обстановки района хранилища и природных процессов. Значительное внимание должно уделяться наличию в геологическом разрезе объектов с аномальными свойствами.
К проектным работам предъявляются требования создания таких систем поверхностного оборудования и буровых скважин, которые обеспечивали бы безопасность и безаварийность их эксплуатации, высокий уровень готовности и ремонтопригодности.
Составной частью программы надежности является программа контроля качества работ, технические условия на прием сооружений и всего подземного хранилища для эксплуатации.
Программой надежности предусматривается организация эффективного контроля за эксплуатацией хранилища. Основной целью контроля является оптимизация процесса захоронения и получение материалов, подтверждающих безопасность эксплуатации хранилища, предъявляемых контролирующим государственным органам.
В состав контроля захоронения входят следующие виды работ:— гидродинамические наблюдения;— гидрогеохимические исследования;— геофизические исследования;— контроль состояния поверхностного оборудования и скважин.По результатам контроля уточняются основные закономерности миграции отхо
дов и протекания сопутствующих процессов, выбирается наиболее благоприятный режим эксплуатации, планируются мероприятия по повышению эффективности и безопасности захоронения.
Оптимизация процесса захоронения и его безопасность может быть обеспечена применением автоматизированных систем контроля и управления, создаваемых на базе ЭВМ.
Автоматизированные системы выполняют следующие функции:— сбор информации контрольных наблюдений о заполнении пласта-коллектора
отходами и протекании сопутствующих процессов, состоянии вышележащих горизонтов и сопредельных площадей;
— обработку и накопление информации;— сопоставление получаемой картины распространения отходов и развития про
цессов в пласте-коллекторе с прогнозными данными;
IAEA-SM-243/110 167 "
— оповещение о повышении вероятности развития аварийных ситуаций;— математическое моделирование режимов эксплуатации полигона захоронения;— принятие решений о проведении мероприятий по обеспечиванию безопасности
захоронения. 'Исходной информацией для автоматизированных систем управления являются
сведения о геологическом строении района захоронения, материалы прогнозов миграции отходов, данные об объеме, давлении нагнетания отходов и их составе, результаты контрольных наблюдений в скважинах, в состав которых входят гидродинамические и геофизические измерения, гидрогеохимические определения.
Применение автоматизированных систем обеспечивает управление процессом захоронения, в том числе процессом распространения отходов в пласте-коллекторе, что значительно уменьшает вероятность возникновения предпосылок для проникновения отходов в биосферу.
Автоматизированные системы позволяют оперативнее получать объективную информацию о состоянии подземного хранилища по первому требованию контролирующих органов.
Выполняемый при осуществлении подземного захоронения комплекс научно-технических работ, позволяет достигнуть необходимой эффективности и безопасности захоронения радиоактивных жидких отходов в глубокие геологические формации.
ЛИТЕРАТУРА
[ 1] СПИЦЫН, В.И. и др., ’’Научное обоснование и практика захоронения радиоактивных жидких отходов в глубокие геологические формации” , Peaceful Uses of Atomic Energy, v. 11 (Proc. 4th Int. Conf., Geneva, 1971) UN, New York, and IAEA, Vienna (1972) 369.
[ 2] С П И Ц Ы Н , в.и. и др.,"Основные предпосылки и практика использования глубоких водоносных горизонтов для захоронения жидких радиоактивных отходов” , Nuclear Power and its Fuel Cycle, v. 4 (Proc. Int. Conf. Salzburg, 1977) IAEA, Vienna (1977) 481.
1 3] ГОЛЬДБЕРГ, В.М., Гидрогеологические Прогнозы Движения Загрязненных Подземных Вод, Недра, М., 1973.
[ 4] БОЧЕВЕР, Ф.М., ОРАДОВСКАЯ, А.Е., Гидрогеологическое Обоснование Защиты Подземных Вод и Водозаборов от Загрязнений, Недра, М., 1972.
[ 5] БЕЛИЦКИЙ, А.С., Охрана Природных Ресурсов при Удалении Промышленных Жидких Отходов в Недра Земли, Недра, М., 1976.
[ 6] ЮДИН, Ф.П., ДОЛГИХ, П.Ф., ВЛАДИМИРОВ, Л.А., Оценка диффузии веществ в глинистых грунтах, Ат. Энерг. 29 5 (1970)5.
[71 СОБОТОВИЧ, Э.В., БОНДАРЕНКО, Г.Н. и др., Изотопно-геохимические Методы Оценки Степени Взаимосвязи Подземных и Поверхностных Вод, Наукова Думка, Киев, 1977.
[ 8] ЛЕМАН, П.Х., ЕЛЬ-БАССИОНИ, А.А.,’’Построение модели надежность - риск для оборудования по удалению радиоактивных отходов” , Доклад №39 США, в кн. ’’Надежность ядерных энергетических установок” , Атомиэдат, М., 1977.
[ 9] РАЦ, М.В.,Структурные Модели в Инженерной Геологии, Недра, М., 1973.[ 10] Справочник по Надежности, Мир, М., 1970.
168 КЕДРОВСКИЙ и др.
DISCUSSION
Н.О. WEEREN: What is the dividing line between high- and intermediate-level
waste (in Ci/litre)? What is the transuranic content of the waste?
M.K. PIMENOV: In this particular case, the intermediate-level waste has a
specific activity of up to 1 Ci/litre; above that we consider the waste as high-level
waste. The transuranics were present in trace amounts, in any case not exceeding
tenths of one mg/litre. This subject is dealt with in more detail in Dr. Balukova’s
paper on the physico-chemical aspects of underground disposal (IAEA-SM-243/113).
J.J.K. DAEMEN: How do you select or determine the injection pressures
and rates? Also, how do you seal boreholes after injection is terminated?
M.K. PIMENOV : In selecting and justifying the spacing pattern for injection
wells and the waste injection conditions the basic requirement is to ensure the
disposal of a given waste volume at the minimum required pressure. Thus, at the
experimental facilities the over-pressures on the formations do not exceed 10—20 atm.
Boreholes which have served their purpose are sealed by methods which are
well known in the oil and gas industry and involve the use of plugging materials
which ensure the isolation of all water-bearing horizons exposed by drilling.
F.A. VAN КОТЕ: Could you indicate below what probability of release
of radioactivity disposal is regarded as acceptable? For what period of time will
the repository be under surveillance?
M.K. PIMENOV: The risk level for disposal is 10_s — 10-6, i.e. the same as
that for nuclear power plants.
Surveillance is planned for the entire period needed for reducing waste activity
to a safe level. Permanent hydrogeological surveillance over the whole territory
of the country is the responsibility of the territorial geological services.
J. HOWIESON: In an article in “New Scientist” Dr. Medvedev claimed that
there had been a serious accident associated with radioactive waste in the Soviet
Union. Can you tell us about this accident or comment on the article?
M.K. PIMENOV: I can state, assuming full responsibility for my words, that
there have been no accident situations in connection with underground disposal
of radioactive wastes. As for Medvedev’s statement, this is a matter between him
and his conscience. If you want details about his publications, you should apply
to him.
P. COHEN : What are the sources of the statistical data on accidents which
you use to study the reliability of the experiments?
M.K. PIMENOV: The statistical data on accidents for calculation and pre
liminary assessment of the reliability of underground disposal of liquid radioactive
wastes are drawn from the experience of the oil industry, as we have had no acci
dents in connection with underground disposal.
P.A. WITHERSPOON: Can you provide some specific details about how you
monitor the migration of radioactive waste away from the emplacement boreholes?
IAEA-SM-243/110 169
M.K. PIMENOV: The migration of radioactive waste is monitored by means
of geophysical, hydrogeological and physico-chemical measurements and analyses.
The results of observations are presented in the form of maps and graphs covering
the whole spectrum of radionuclides.
A. AZIZ: Do you carry out seismic surveys around the disposal site after
the introduction of wastes in order to observe any tectonic activity of the earth
incidental to the introduction of these wastes?
M.K. PIMENOV: The sites for experimental work on underground disposal
of liquid wastes are selected only in an area of low seismic activity (not more than
6—7 points on the Richter scale). Seismic monitors have been installed at the
experimental sites recently in order to study the seismic effect of pumping waste
into the formations. As measurement results are accumulated, they will be published.
K. ARAKI: What is the depth of boreholes for liquid waste injection? What
is the rate of injection in m3/day?
M.K. PIMENOV: The depths of the boreholes at the experimental sites are
350—500 m and 1500 m and the injection rates are 700—1000 m3/day and
300—350 m3/day, respectively.
IAEA-SM-243/42
WASTE DISPOSAL BY SHALE FRACTURING
AT OAK RIDGE NATIONAL LABORATORY
H.O. WEEREN
Oak Ridge National Laboratory,
Oak Ridge, Tennessee,
United States of America
Abstract
WASTE DISPOSAL BY SHALE FRACTURING AT OAK RIDGE NATIONAL LABORATORY.The shale fracturing process is a method of waste disposal currently in use at Oak Ridge
National Laboratory for the permanent disposal of certain locally generated radioactive waste solutions. In this process, the waste solution is mixed with a solids blend of cement and other additives; the resulting grout is then injected into an impermeable shale formation at a depth of 200 to 300 m. The grout sets a few hours after completion of the injection, fixing the radioactive waste in the shale formation. Operational experience with this process is discussed.A description of a new facility being built and the preliminary-site proof test that was required is given.
Shale fracturing is a process currently being used at Oak Ridge National Laboratory (ORNL) for the permanent disposal of locally generated intermediate-level waste solutions. These solutions are alkaline, about 1 M in NaNOq, and have a radionuclide content (predominantly i37Cs) of about 0.2 Ci/ltr. In this process, the waste is mixed with a solids blend of cement and other additives; the resulting grout is then injected into an impermeable shale formation at a depth of 200 to 300 m, well below the level at which groundwater is encountered. During the course of the injection, the injected grout forms a thin, approximately horizontal grout sheet 100 to 200 m in width. The grout sets a few hours after completion of the injection, permanently fixing the radioactive waste in the shale formation.
The essential feature of the shale fracturing process is the fixation of the radionuclides in a geological formation that is known to be isolated from contact with the surface environment. The process has additional features that would provide continued containment of the radionuclides even if the isolation of the disposal formation should be lost. For example, the leach rates of significant radionuclides from the set grout are quite low. In addition, any radionuclides that might be leached from a grout sheet would be retained in the disposal zone by the high ion-exchange capacity of the shale; therefore, this process offers an exceptionally favorable approach to permanent disposal of radioactive wastes.
171
172 WEEREN
DRV SOLIDS STO RAGE BINS
PUMP HOUSE
VA LV E PIT
EM ERGENCY WASTE TRENCH
WASTE STO RAGE TAN KS
GR A Y SHALE
LIMESTONE BED
CASED
O BSERVAT IO N WELL
GROUT SHEETS
FIG. I. ORNL fracturing disposal pilot plant process schematic.
The mix developed for this process consists of Portland cement, fly ash, drilling clay, pottery clay, and a retarder. The retarder delays the setting time of the mix, the pottery clay fixes cesium, the drilling clay retains excess water, the fly ash fixes strontium, and the cement is the overall binder. These various solids are blended and stored just before each injection. This blend is subsequently mixed with the waste solution in a ratio of about 0.9 kg of solids per liter of waste solution. The resulting grout has a density of about 1.5 g/cm3 and an apparent viscosity of about 40 cP. The grout remains fluid for about 24 h, if kept in motion. The compressive strength of the set grout is low (about 1.5 MPa). The rates at which radionuclides can be leached from the set grout are also quite low (i.e., approximately equivalent to those from a borosili- cate glass). The cesium leach rate is 7 yg/(cm2-d), the strontium is 32 yg/(cm2,d), and curium and plutonium are about 0.15 ug/(cm2-d). These rates were determined for specimens aged 28 and 100 d [1].
Each injection disposes of an annual accumulation of waste solution of about 300 000 Itr. Prior to the injection, the waste solution is pumped to the waste storage tanks at the injection site. The dry solids are blended and stored in bins at the injection facility. A standby injection pump is rented for each injection; its function is
IAEA-SM-243/42 173
to clean grout from the injection well in the event of failure of the main injection pump. During the injection, the waste solution is pumped to the mixer, continuously mixed with the preblended solids, and then discharged into the surge tank. From the surge tank, the grout is pumped down the tubing string in the injection well and out into the shale formation. A schematic of the process is shown in Fig. 1. The injection pressure is about 200 atm. The normal grout injection rate is about 1000 ltr/min; an injection requires about 8 h to complete. The grout sheet formed during the injection is approximately 1 cm thick and up to 200 m wide. The fracture orientation generally follows the bedding planes in the shale, which are inclined about 10 to 15° to the horizontal. At the end of the injection, the well is flushed with water so that the slot in the injection well will be free of grout and can be reused for the next injection. Then a valve shuts the well until the grout has set. Subsequent injections are made through the same slot, forming grout sheets that are generally parallel to the first. After four injections have been made through the one slot, the bottom of the well is plugged and a new slot is cut in the casing of the well 3 m above the old slot. The surrounding shale formation is fractured at this new depth by pressurizing the well until a sudden drop in pressure signals the creation of a fracture.
The radiation exposure of the operating crew and the injection pressure are regularly monitored during each injection. A few days after the injection, the orientation of the grout sheet is determined by logging the network of observation wells that surrounds the facility. (These are cased wells that extend to the bottom of the disposal formation.) A gamma-sensitive probe lowered in these wells detects the presence of the grout sheet at a particular depth, thereby verifying the orientation of the grout sheet. A representative series of logs is shown in Fig. 2. After several injections have been completed, the cumulative surface uplift around the injection well is determined by measuring the change in elevation of a network of bench marks.This uplift averages 0.03 cm per injection at the injection well and decreases regularly to near zero at about 400 m from the well. The significance of this measurement is dubious, and it will probably be discontinued. The permeability of the shale overlying the disposal zone is also periodically measured to verify that it has not been increased by the stresses generated by repeated injections. No change in the cover rock permeability has been observed to date.
The process was developed in a series of experiments between 1959 and 1965. The experimental facility was modified in 1966 for the routine disposal of intermediate-level waste solutions generated at ORNL. Since 1966, this facility has been used for 17 operational injections. More than 8 Mltr of waste grout containing over 600 000 Ci of radionuclides have been injected. Although operational problems have been experienced, most have been comparatively minor and none has been severe; the general experience has been quite good. With the exception of four injections (discussed below), the difficulties have not been serious enough to force the termination or major delay of an injection; they have required, at most, a relatively short shutdown of the injection while repairs were being made. These difficulties included (1) eroded check valves in the injection pump, (2) a
174 WEEREN
о
- 6
i -12
ш -зо
-3 6
FIG.2. Representative series of logs of grout sheet.
plugged drain line from the injection pump sump, (3) a ruptured solids supply-line connection, (4) loss of prime in the waste pump, (5) jamming of the clutch on the injection pump, (6) bridging of solids in the feed hopper, and (7) a leak past the sealing ring in one of the high-pressure valves. Each incident was an isolated occurrence and none caused serious difficulty.
One delayed injection resulted from a failure of a packing seal in the injection pump. In this case, the facility and well were washed free of grout with the standby pump; repairs were made, and the injection was resumed 2 d later. In another injection, the drain valves on the high-pressure valve rack were eroded by leakage of grout through the valves. The valves would no longer hold pressure; therefore, the injection was halted, the facility and well were washed free of grout, repairs were made, and the injection was resumed 2 d later.
One injection that was terminated resulted from an attempt to use blended solids that had been stored for several months. The flowability of these solids was poor, and the injection was quickly shut down. Another injection was terminated when the diesel drive of the injection pump threw a connecting rod through the block. The facility and well were washed with the standby pump.
General experience with the shale fracturing facility in 7 experimental and 17 operating injections has been quite good. Large volumes of waste solution have been continuously mixed with dry solids, in the desired proportions, and injected into the isolated shale bed. Cleanup of small waste spills is feasible, as is the direct maintenance of mechanical equipment.
IAEA-SM-243/42 175
The operational cost of an injection is approximately $50 000 (US). About $10 000 of this is the cost of the dry solids, about $25 000 is the service charge of an oil well cementing company for making the injection, and the remaining $15 000 is for various maintenance and operations charges.
A new shale fracturing facility is being designed and built at a site about 250 m south of the existing facility. At this location, the disposal zone is about 60 m deeper than at the existing facility, the geology is similar in other respects. A site proof test was made at the new site to verify that the site was suitable for waste disposal by shale fracturing. This test consisted of drilling an injection well and four observation wells at the site and making a test injection of grout tagged with a radioactive tracer. The injected grout was detected in three of the observation wells at depths that indicate that bedding-plane fractures were formed. Subsequently, a water injection was made to obtain pressure decay data. This test indicated that no extensive interconnected fractures and joints exist at the disposal site and that the shale permeability is very low at the injection depth. The results of these tests indicate that the site is suitable for shale fracturing disposal operations.
An environmental impact statement has been written to cover the operations of the facility [2]. The statement concludes that the overall impact would be beneficial. The facility would remove large volumes of potentially hazardous radioactive wastes from the existing surface storage facilities and would fix these wastes in impermeable shale formations (well removed from the biosphere). All major incident situations postulated are considered quite improbable, and the analysis of each case indicates that the ultimate release of radionuclides to the environment would be small.
The new facility will have improved shielding and containment so that wastes of higher specific activity can be handled. These wastes are expected to include currently generated intermediate- level wastes, resuspended sludges that have accumulated in waste storage tanks over the past thirty-five years, and pilot plant wastes with a specific activity of up to 8 Ci/ltr. Very little of thiS'latter waste is expected, but it was made the design-basis waste for the new facility. The operating pressures and flow rates for the new facility will be similar to those of the existing facility. The dry-solids handling equipment, which has been a source of chronic difficulty in the existing facility, will be improved so that the flow of solids to the mixer will be smooth and controlled. The process instrumentation will be improved by the incorporation of a weigh- belt feeder to measure the flow of solids more precisely. Improved mix ratio indicators will be installed to determine and display the ratio of the weight of solids and the volume of liquid going to the mixer. This ratio should be kept within rather narrow limits for good process control. A check on this ratio will be provided by the ratio of grout volume to liquid volume, a ratio that is directly proportional to the mix ratio. Completion of construction of the new facility is scheduled for early 1981, and the first injections will be made at the new facility shortly thereafter.
176 WEEREN
REFERENCES
[1] MOORE, J. G., et al., Development of Cementitious Grouts for the Incorporation of Radioactive Wastes, Oak Ridge National Laboratory Rep. ORNL-4962 (1965).
[2] Final Environmental Impact Statement, Management of Intermediate Level Radioactive Waste, ERDA-1553 (1977).
DISCUSSION
H. KRAUSE: I have no doubts about the safety of hydraulic fracturing
under the particular conditions prevailing at the site you have chosen. However,
it seems to me that this process results in a very large surface with the waste. I
should therefore like to ask if the red shale underlying Oak Ridge has sufficient
mechanical stability for the excavation of caverns. Have you ever considered the
possibility of pouring the grout into such caverns, thus producing a large mono
lithic block with a favourable volume-to-surface ratio?
H.O. WEEREN: This possibility has not been seriously considered; the
performance and safety of the shale fracturing facility has been so satisfactory
that there has been little incentive to look at alternatives. The high surface-to-
volume ratio has compensating advantages; heat dissipation is much more rapid
from grout sheets than from a monolith, for instance.
R. KÔSTER: Do you have any information on the Sr retention effects
for different fly ash qualities?
H.O. WEEREN: We have used two different sources of fly ash and have
seen no difference with respect to Sr retention. No serious study of this particular
variable has been made.
R. KÔSTER: You give the leach rates of Cs and Sr for your products and
indicate that the rate for Cs is lower. In the case of normal cements without,
additive the leach rate of Cs is about 10 times as high as that of Sr. This is con
sistent with our results from Karlsruhe — it is easier to find an additive for Cs
retention than for Sr retention.
K. KÜHN: The process of hydraulic fracturing was developed specifically
for the geological conditions at the Oak Ridge site and for the site-specific opera
tions at ORNL. Do you think the process can be applied to other types of waste
and to other geologic sites? Some years ago, for example, there were discussions
about using the process for the disposal of the high-level waste being held at the
NFS site at West Valley, N.Y. Has any decision been reached so far?
H.O. WEEREN: The shale fracturing process is probably applicable to most
alkaline waste solutions, possibly excluding those wastes containing boron. The
composition of waste solutions has changed considerably over a period of years
IAEA-SM-243/42 177
and no difficulty has been experienced for this reason. The process is, of course,
quite dependent on site geology.
Discussion of the best method for handling the wastes at West Valley is still
continuing; no decision has yet been taken.
M.K. PIMENOV: High pressures are needed for hydraulic fracturing,
especially in the initial period. What equipment do you use to produce the
pressure for hydraulic fracturing - standard oil-well equipment or equipment
designed specially for pumping radioactive wastes with due provision for
radiation safety?
H.O. WEEREN: The injection pump is a standard oil-field cementing pump,
manufactured and used by Halliburton Co. This pump is installed in a cell so that
any waste escaping from the pump will be retained and will not escape into the
environment.
H.M. HARSVELDT: Assuming the shale to be very fissile, are you not
afraid of creating gliding planes by the injection of grout sheets and thus of causing
earthquakes or landslides, as a result of which the grout sheets would come into
contact with the biosphere? Is the area regularly monitored?
H.O. WEEREN: The question of possible generation of earthquakes has been
considered because of the experience at Denver (where deep-well injection did
generate earthquakes). It was concluded that the shale fracturing process is only
superficially similar; one major difference is that, while at Denver liquid waste
was being injected, at Oak Ridge we inject a grout. The area is monitored for
seismic activity; none attributable to the injections has been observed.
L.J. ANDERSEN: What is the total thickness of the injected waste material?
Have you observed hydraulic fractures outside the grouted area and any artificial
increase in permeability around the waste?
H.O. WEEREN: The total thickness after 18 injections should be about
18 cm (it has not been cored and measured, but several measurements for individual
injections indicate an average thickness per injection of 1 cm).
The maximum extension of grout sheets from the injection well that has
been observed is 700 ft. The maximum extent of observed radionuclide movement
is the same. There is no evidence to show that the radionuclides migrate beyond
the grout sheet. The permeability is not measured in the disposal zone but several
hundred feet above it. No increase in permeability in this zone has been observed.
R. KRAEMER: Can you give an indication of the specific costs per cubic
metre of injected waste, including capital and operational costs?
H.O. WEEREN: These costs would depend largely on the frequency of
injection and the volume of waste injected. The capital cost of the new facility
would be about US $5 to 6 million; the operating cost for a single injection of
100 000 gal would be about $50 000.
IAEA-SM-2*3/113
ФИЗИКО-ХИМИЧЕСКИЕ ПРОЦЕССЫ ПРИ УДАЛЕНИИ
ЖИДКИХ РАДИОАКТИВНЫХ ОТХОДОВ
В ГЛУБОКИЕ ПЛАСТЫ-КОЛЛЕКТОРЫ
В.И.СПИЦЫН, В.Д.БАЛУКОВА
Государственный комитет
по использованию атомной энергии,
Москва,
Союз Советских Социалистических Республик
Abstract- Аннотация
PHYSICAL AND CHEMICAL PROCESSES OCCURRING AS A RESULT OF THE DISPOSAL OF LIQUID RADIOACTIVE WASTES IN DEEP FORMATIONS.
A number of physical and chemical processes requiring evaluation, which occur as a result of the disposal of liquid radioactive wastes within deep formations, are discussed. The main requirements to which the use of such formations for waste disposal are subject are reliable hydrogeological conditions and non-disturbance of the natural structure of the formation throughout the period required for the wastes to become harmless. Physical and chemical studies on the following subjects are discussed in connection with the second of the above requirements: ( 1) the chemical interaction of wastes with the formation material; the conditions under which individual minerals are broken down; the reactions of the formation water with wastes and the interstitial sedimentation; (2) the interphase distribution of macro- and micro-components in formation repositories occurring as a result of hydrolysis of salts and precipitation of surface-active matter; (3) laboratory and field experiments for determining sorption and migration of radionuclides, with an evaluation of the maximum accumulation of radionuclides in the solid phase of formations as a function of the radioactivity of the wastes, the types of nuclide and the filtration parameters; (4) examination of the adsorption characteristics of multimineral aluminosilicate systems in relation to various radionuclides, with a physical and chemical evaluation of the active sorption centres of mineral surfaces (infra-red spectra and diffraction patterns); (5) experimental evaluation of the chemical treatment of individual rocks and other minerals in order to control the adsorption parameters and the accumulation of radioactivity in the solid phase of the formation, the extent to which previous chemical activation influences nuclide accumulation; ways in which adsorption capacity can be lowered by treating rocks with surface-active and complex-forming reagents;(6) the mechanism and firmness with which various nuclides become fixed to rocks and the results of desorption experiments; (7) the effects of ionizing radiation on rocks; evaluation of rock stability; extent and nature of changes in sorption parameters of aluminosilicates induced by irradiation with gamma quanta and electrons; and (8) evaluation of transuranium element behaviour within formations.
179
180 СПИЦЫН и БАЛУКОВА
ФИЗИКО-ХИМИЧЕСКИЕ ПРОЦЕССЫ ПРИ УДАЛЕНИИ ЖИДКИХ РАДИОАКТИВНЫХ ОТХОДОВ В ГЛУБОКИЕ ПЛАСТЫ-КОЛЛЕКТОРЫ.
Рассматривается комплекс физико-химических процессов, требующих оценки, при удалении жидких радиоактивных отходов в глубокие пласты-коллекторы. Основополагающими принципами использования таких пластов для хранения отходов являются надежные гидрогеологические условия и осуществление захоронения без нарушения естественной структуры формации, в течение всего времени обезвреживания отходов. Для обоснования второго положения рассматриваются результаты следующих физико-химических исследований. 1. Химическое взаимодействие отходов с пластовым материалом, условия разрушения отдельных минералов, реакции пластовой воды с отходами, осадкообразование в поровом пространстве. 2. Межфазное распределение ма- кро- и микрокомпонентов в условиях пластов-хранилищ, являющееся следствием гидролиза солей и осаждения поверхностно-активных веществ. 3. Лабораторные и полевые эксперименты по определению сорбции и миграции радионуклидов, с оценкой максимального накопления радионуклидов на твердой фазе пластов в зависимости от уровня радиоактивности отходов, от форм существования нуклидов и фильтрационных параметров. 4. Адсорбционные характеристики многоминеральных алюмосиликатных систем по отношению к различным радионуклидам. Физико-химическая оценка активных сорбционных центров поверхности минералов (ИК-спек- тры и дифрактограммы). 5. Экспериментальная оценка химической обработки отдельных минералов и пород, проводимой с целью регулирования адсорбционных параметров и накопления радиоактивности на твердой фазе пластов. Степень влияния на накопление нуклидов предварительной химической активации. Условия снижения адсорбционной способности вследствие обработки пород поверхностно-активными и комплексообразующими реагентами. 6. Механизм и прочность закрепления различных радионуклидов на минералах пород. Результаты экспериментов по десорбции. 7. Результаты по воздействию ионизирующего излучения на породы, оценка уровня их стабильности. Степень и направление изменений сорбционных параметров алюмосиликатных минералов вследствие облучения 7 -квантами и электронами. 8. Оценка поведения трансурановых элементов в пластах-коллекторах.
1. ВВЕДЕНИЕ
Жидкие радиоактивные отходы атомных предприятий представляют собой слож
ные неустойчивые системы. В зависимости от происхождения жидких отходов неустой
чивость системы определяется либо наличием легкогидролизующихся ионов металлов,
либо присутствием органических и поверхностно-активных веществ, либо физико-хи-
мическими процессами, связанными с изменением форм существования макросостав
ляющих растворов. При захоронении в глубокие пласты-коллекторы жидкие отходы
удаляются в не менее сложную по свойствам геохимическую среду, представленную
гетерогенной системой минералов и природными растворами — подземными водами,
основные компоненты которых являются комплексными соединениями и простейшие
из них—аквакомплексы. При удалении жидких радиоактивных отходов происходит
нарушение физико-химических равновесий как в системе отходов, так и в геохимичес
кой среде формации, т.е.изменяется межфазное распределение компонентов, а также
энергетические характеристики геологической среды. Все это требует детальной оценки
перед началом использования хранилищ, и настоящее сообщение посвящено физико
химическим аспектам, связанным с поведением отходов при их удалении в пласты-кол-
лекторы.
IAEA-SM-243/113 181
2. ХИМИЧЕСКОЕ ВЗАИМОДЕЙСТВИЕ
Основной принцип использования природной геологической формации для хра
нения жидких отходов — это надежные гидрогеологические условия и сохранение их,
а следовательно, и структуры пластов, практически неизменными на весь период обез
вреживания отходов. Поэтому система отходов не должна обладать свойствами, нару
шающими естественную структуру пласта-коллектора, т.е., с химической точки зрения,
отходы не должны разрушать минералы, их естественную связность. Соблюдение тако
го принципа исключает удаление агрессивных по отношению к минералам пород жид
костей и требует проведения технологической подготовки отходов до такого состояния,
чтобы их взаимодействие с пластом не приводило к разрушению пород.
Вместе с тем, в составе отходов имеется много компонентов, образующих в ус
ловиях пласта осадки (ионы легкогидролирующихся металлов) или сорбирующихся
на минералах пород (поверхностно-активные вещества), тем самым препятствующих
использованию полезного объема хранилища. Этим положением определяется необ
ходимость предварительной стабилизации отходов с целью исключения или замедле
ния осадкообразования в пласте, что также достигается технологической подготовкой
отходов перед их захоронением [1,2].
3. МЕЖФАЗНОЕ РАСПРЕДЕЛЕНИЕ КОМПОНЕНТОВ ОТХОДОВ
Осадкообразование в поровом пространстве в значительной степени влияет на
межфазное распределение радионуклидов, увеличивая удельную долю активности в
твердой фазе.
Породы пластов-коллекторов представлены главным образом алюмосиликатны-
ми или карбонатными (доломиты и известняки) минералами. Кислотность среды при
родных пластовых вод находится, как правило, в пределах рН = 6,5-7,8, поэтому одним
из наиболее легко реализуемых в пластовых условиях процессов является гидролиз
металлионов — продуктов коррозии, по современным представлениям находящихся в
состоянии аквакомплексов.
При этом образование осадков гидроокисей, а также скорость их старения, а сле
довательно и структура, зависят от температурных условий, а кинетика разложения
жидкой фазы и ряд обратимых реакций — от температуры и давления.
Проведенная оценка межфазного перераспределения радиоизотопов при образо
вании осадков в поровом пространстве показала, что гидролиз солей металлов при
водит к практически полному удалению из жидкой фазы всех коллоидных, псевдо-
коллоидных и полимерных форм радионуклидов. Данные формы являются зародыша
ми, на поверхности которых наращиваются гидроокисные образования, которые в свою
очередь являются сорбентами для других форм радионуклидов. Чем медленнее про
ходит процесс гидролиза, тем больше работающая на сорбцию поверхность и выше
удельная активность гидроокисей. При быстром течении процесса, стимулируемого
182 СПИЦЫН и БАЛУКОВА
ТАБЛИЦА I. ВЛИЯНИЕ УСЛОВИЙ ГИДРОЛИЗА
НА ВЫВЕДЕНИЕ РАЦИОСТРОНЦИЯ ИЗ ЖИДКОЙ
ФАЗЫ (количество удаленного 90Sr в %)
Компонент,подвергающийсягидролизу
Температура гидролиза
иООTill t = 140° С
Fe3+ 98 до 1,2
Сг3* 92 0,5
температурным разогревом, получается более плотная структура гидроокисных осад
ков, но при этом резко снижается сорбционное удаление радионуклидов. В табл.1 пред
ставлены сравнительные значения сорбционного удаления радиостронция (ионная фор
ма) в зависимости от условий гидролиза. Вторичный вывод радиоизотопов в жидкую
фазу с таких осадков без растворения последних возможен лишь ограниченно и не от
носится к полимерным, коллоидным и псевдоколлоидным формам, которые доста
точно надежно перекрыты от воздействия на них жидкой фазы.
Инфракрасные спектры (ИК-спектры) гидроокисных осадков, выпавших в ус
ловиях высоких температур, показывают, что в начальной стадии образования осад
ков фиксируются оловые связи, которые очень быстро исчезают, и выявляется осно
вная связь металл-группа ОН. Естественно, что в последней стадии старения гидро
окисей накладывается фактор индивидуальности структуры гидроксидов. Число .
остающихся активных центров при старении резко снижается. Следовательно, эффек
тивность удаления из жидкой фазы радионуклидов при гидролизе зависит от природы
осадков, а также от условий гидролиза, при прочих равных условиях, таких, как кон
центрация нуклида, ионная сила и др.
Второй причиной нарушения стабильности жидкой фазы в пласте являются ре
акции взаимодействия высокоминерализованных пластовых вод с отходами, содер
жащими сульфаты, карбонаты и другие анионы и поверхностно-активные вещества (ПАВ) ,
дающие осадки с катионами пластовых вод. Для отходов характерны неионогенные и
анионоактивные ПАВ. Как правило, такие отходы являются мицеллярными структура
ми. В условиях пласта они коагулируют в осадки и сорбируются на поверхности мине
ралов. Их влияние на межфазное распределение радиоизотопов может быть большим,
поскольку кинетически процесс сорбции ПАВ на минералах проходит быстрее, чем ионо
обменные реакции радионуклидов. Сорбционные емкости пород по отношению к ПАВ
достаточно велики и поэтому они постепенно поглощаются из жидкой фазы. Слои ПАВ,
образовавшиеся на поверхностях минералов, меняют их молекулярную природу и свой
ства: может меняться заряд поверхности и происходит экранирование активных цент
ров. Этим можно пользоваться для регулирования накопления радионуклидов на по
верхности твердой фазы. Коагулирующие в осадки ПАВ захватывают радионуклиды,
IAEA-SM-243/113 183
ТАБЛИЦА П. СРЕДНЕЕ ЗНАЧЕНИЕ КОЭФФИЦИЕНТОВ
РАСПРЕДЕЛЕНИЯ
Раствор ПородаКоэффициент распределения (КР)
КР (,0Sr ) КР (137Кг )
Песчаник 21,7 295,5
Фоновый Алевролит 14,9 43,0
Известняк 7,3 9,0
Песчаник 1 , 6 98Дезакти-вационный
Алевролит 1 , 0 23
Известняк 1,5 5
а сорбирующиеся на поверхностях минералов препятствуют сорбции нуклидов. Таким об
разом, с помощью ПАВ, меняя их природу и количественное содержание, можно прак
тически исключать и увеличивать накопление радионуклидов в твердой фазе, а в опре
деленных случаях усиливать или уменьшать их десорбируемость. В табл.II приведены
примеры снижения коэффициентов распределения радионуклидов при обработке по
род растворами, содержащими ПАВ.
Влияние ПАВ можно оценить следующими величинами: как правило, в 5-6 раз
увеличивается сорбционное накопление в присутствии катионоактивных веществ, в
1 ,2 -1 ,6 раз — для сульфанолов и алкилсульфатов, карбоновых кислот и нефтяных суль
фокислот. В присутствии анионоактивных веществ закрепление нуклидов на породах
непрочное. Комплексующие реагенты, дающие водородные связи с гидроксильными
группами твердой фазы, дают положительный эффект сорбции и регулируют леофиль-
ность поверхности. С точки зрения оценки безопасности, указанные физико-химические
процессы межфазного распределения компонентов отходов являются дополнительным
к гидрогеологическим условиям барьером, обеспечивающим локализацию радионукли
дов в ограниченных пределах формации.
4. СОРБЦИЯ И МИГРАЦИЯ РАДИОНУКЛИДОВ С ФИЛЬТРУЮЩИМСЯ ПОТОКОМ
Введение радионуклидов в пласты-коллекторы и их межфазное перераспределе
ние в конечном счете меняет энергетические характеристики в геологической среде.
Последнее дает реально ощутимые результаты для отходов с уровнем активности,
превышающим 10“б Ки/л, когда заметно изменяется (на 2-3 порядка) удельное содер
жание радионуклидов в пласте. Одновременно, при таких уровнях активности в ус
ловиях пласта-хранилища под.влиянием процессов радиолиза макрокомпонентов и
1 8 4 СПИЦЫН и БАЛУКОВА
ТАБЛИЦА III. МАКСИМАЛЬНО ВОЗМОЖНОЕ НАКОПЛЕНИЕ СТРОНЦИЯ И
ЦЕЗИЯ (мг/100г)
Состав раствораКаолинит
Sr Cs
Хлорит
Sr Cs
NaNOj - 200г/л, NaOH до 4 г/л. Обработка пластовой водой.
Предварительная обработка H N 0 3 - 15 г/л
Предварительная обработка NaOH - 10 г/л
Грунт облучен (7 -облучение, в°Со, 8,3 • 1 0 ' рад) и обработан H N 03 —15 г/л
Грунт облучен (7 -облучение, í 0 Co, 8 ,3 -10“ рад) и обработан NaOH
3
2.7
4
4,5
5.7
1,5
3,1
5,9
6,0
6,2
2,1
1,9
примесей веществ в коллоидном состоянии, а также в результате ряда реакций с плас
том, происходит трансформация форм радионуклидов. Концентрирование радиону
клидов на твердой фазе осуществляется в небольшом объеме пласта достаточно быс
тро, с понижением уровней активности жидкости до Ю^-КГ7 Ки/л. Вследствие этого
основные объемы жидких отходов в пласте-коллекторе содержат лишь незначитель
ные количества активности по сравнению с исходным уровнем и переходят в катего
рию низкоактивных жидкостей. Так, при исходных активностях 10"3-10 Ки/л в
жидкой фазе остаточная радиоактивность в пласте составляет от 1СГ5 до 10~2% от ис
ходной. Процессы десорбции с твердой фазы для большинства компонентов и форм
радионуклидов в фильтрующемся потоке отходов практически неосуществимы, а
для тех форм, которые могут быть десорбированы, кинетика процессов замедлена
не менее чем на порядок по сравнению с кинетикой сорбции. Поэтому смещение фрон
та жидкости уже на уровне активности 1СГ5-1СГ6 Ки/л происходит со значительным
замедлением по сравнению с движением непосредственно жидкой фазы [3].
Если исключить влияние гидролиза многовалентных катионов металлов и адсорб
ционный вывод макрокомпонентов ПАВ из жидкой фазы, то количественные пока
затели распределения радионуклидов при контакте с пластовыми минералами подчи
няются закономерностям ионного обмена с учетом конкурирующих реакций обмена
макроионов. Имеют значение среда (pH), ионная сила и форма, в которой находится
радиоизотоп, а также адсорбционная способность минералов породы. Общие законо-
IAEA-SM-243/113 1 85
ТАБЛИЦА IV. ДИНАМИЧЕСКАЯ СОРБЦИОННАЯ ЕМКОСТЬ
(исходная активность — (4-7) ■ 10~6 Ки/л)
Скорость Сорбционная
Изотоп Порода Раствор фильтрации(м/сут)
емкость (10-‘ Ки/л)
Песчаник 0,57 52,0
Алевролит Фоновый 1,3 14,5
Известняк 1,15 6,7
90Sr
Песчаник 1,31 2,9
Алевролит
Известняк
Дезактива-ционный
1,28
1,40
0,8
3,5
Песчаник 1,12 100,5
Алевролит Фоновый 1,28 89,7
Известняк 1,40 30l37Cs
Песчаник 1,12 6,7
Алевролит
Известняк
Дезактива-ционный
1,03
1,15
3,0
1,6
мерности включают: эквивалентный обмен ионов ; поглощение катионов (тем боль
ше, чем выше валентность) ; увеличение поглощения эквивалентно возрастанию концен
траций; величина и скорость обмена зависят также от природы обменивающихся ионов,
структуры адсорбента, времени взаимодействия и температуры.
Для реальных минералов пород, на которых сорбируются радионуклиды, ИК-спек-
троскопические и дифракционные исследования позволили установить наличие полос
поглощения и рефлексов, характерных для силоксановых и силонольных групп тетра-
и октаэдрического алюминия и групп ОН. Адсорбционная способность минералов
обусловлена как наличием обменных катионов, так и гидроксильными группами по
верхности. Общая обменная емкость может быть оценена потенциометрическим ме
тодом.
Большое влияние на сорбцию оказывает химическая активация минералов. В
табл.Ш показано влияние химической активации пород и облучения на поглощение
радионуклидов. Как видно, сорбционные характеристики природных глин при таких
обработках повышаются. Кислотная активация приводит к замене обменного комп
лекса на ион Н+, щелочная активация — к постепенному изменению на поверхности
186 СПИЦЫН И БАЛУКОВА
ТАБЛИЦА V. ДИНАМИЧЕСКАЯ СОРБЦИОННАЯ ЕМКОСТЬ
(исходная активность 2 -10-4 Ки/кг)
Изотоп ПородаСорбционная емкость
(Ки/л)
s°SrМергель ' 2,4 ■ 10-3
Глина 1 ,16-Ю -3
,3,CsМергель 13,8-10-3
Глина 8 -К Г 3
структуры кремнеалюмокислородныхтетраэдров и октаэдров. Ионизирующее излу
чение в совокупности с химической активацией усиливает разупорядоченность струк
туры. Происходит не только увеличение сорбционной емкости, но и прочное закре
пление радионуклидов в твердой фазе.
Как уже было указано, большое значение для количественных показателей ад
сорбции (коэффициент распределения и сорбционная емкость) имеют абсолютная кон
центрация радионуклидов и доля радиоизотопа в ионной силе отходов, т.е. влияние
конкуренции макросостава отходов. Если для отходов среднего и высокого уровней
активности возможно достаточно большое накопление радионуклидов на твердой фа
зе при значительных коэффициентах очистки отходов, то в низкоактивных отходах,
особенно если в них присутствуют ПАВ, коэффициенты распределения значительно
ниже и практическое накопление радионуклидов в формах, способных к миграции,
также существенно ниже максимальных величин. В качестве примера представлены
данные в табл.IV и V. Сравнивая их, можно видеть, что увеличение исходного содер
жания радионуклидов на два порядка приводит к увеличению сорбционной динами
ческой емкости, т.е. к накоплению на 2-3 порядка.
В среднем для всего диапазона радиоактивности низкоактивных отходов кон
центрирование радионуклидов (за исключением трития) происходит не менее чем на
порядок.
Концентрация 90 Sr возрастает в 50 раз, 137Cs — в 100-600 раз, 144Се, в зависимости
от форм существования, — в 1 0 0 и более раз.
Миграция радионуклидов в пласте-коллекторе при плоскопараллельном потоке
подземных вод определяется с учетом трех одновременно протекающих процессов:
фильтрации, гидравлической дисперсии и сорбции. Все три процесса в общем случае
описываются дифференциальным уравнением
IAEA-SM-243/113 187
где С — концентрация изотопа; N — количество сорбированного вещества, отнесен
ное к единице объема пор среды; п — активная пористость пород (отношение откры
тых пор к объему породы) ; U - скорость фильтрации; D — коэффициент гидравли
ческой дисперсии (фильтрационная диффузия) ; t — время.
Уравнение баланса (1) решается совместно с уравнением кинетики сорбции, ко
торое для небольших концентраций сорбируемого вещества в ионной форме имеет
вид
. | ^ = a (C - 0 N ) (2 )
где а — константа скорости сорбции; 0 — коэффициент распределения радиоизотопов
между жидкой и твердой фазами, равный Co/No, где С0 и N0 — концентрации изотопа
в жидкой и твердой фазах, соответственно.
Кинетикой сорбции микроконцентраций ионов в условиях пласта можно прене
бречь и, принимая условия мгновенного установления равновесия, при прогнозах дви
жения радиоизотопов можно пользоваться уравнением изотермы сорбции.
Указанные уравнения решаются для граничных условий динамики сорбции, и
все решения их требуют знания следующих параметров: активной пористости среды(п0) ;
коэффициента фильтрационной диффузии (D); коэффициента распределения (|3) и
константы скорости сорбции ( а ) .
Эти параметры зависят от свойств отходов и пород пласта и определяются экс
периментально в каждом случае путем динамических опытов на керновом материале.
Результаты расчетов и экспериментов показали, что отставание нуклидов от фрон
та движущегося потока является значительным и для каждого изотопа индивидуальным.
Так, для стронция получаются коэффициенты от 3 до 60, а для цезия — от 10 до 100.
Сопоставление экспериментальных данных для различных форм нуклидов в от
ходах позволило установить причины миграции отдельных радионуклидов. Таких при
чин три: наличие в отходах поверхностно-активных веществ, наличие в отходах комп
лексных соединений, определенное равновесное межфазное распределение всех ком
понентов в пластовых условиях [1]. Вместе с тем экспериментально доказано, что в
пластах-коллекторах, даже если выход загрязнений на уровне предельно допустимых
концентраций недалек от фронта движения жадкости, увеличение уровня ми
грирующей активности происходит очень медленно, при постоянном накоплении ну
клидов на твердой фазе.
5. ВЛИЯНИЕ ОБЛУЧЕНИЯ НА МИНЕРАЛЫ ПОРОД, СОСТАВЛЯЮЩИХ
ГЕОЛОГИЧЕСКУЮ СРЕДУ ХРАНИЛИЩ ОТХОДОВ
Известно, что радиационная устойчивость неорганических сорбентов типа окис
лов достаточно высокая, вплоть до доз 101Орад [4]. Вместе с тем очевидно, что и в
устойчивых структурах при большом накоплении микродефектов могут изменяться
188 СПИЦЫН и БАЛУКОВА
поверхностные свойства, например, сорбционные параметры: кинетические зависимо
сти и возможная степень накопления вещества, т.е. способность удерживать или обме
нивать токсичные компоненты.
Характер радиационных повреждений под воздействием 7 -квантов и электрон
ного потока в минералах алюмосиликатного состава зависит как от свойств облучае
мой среды (характер связи, энергия кристаллической решетки и др.) , так и от ряда
условий облучения (температура; среда, в которой проводится облучение; присутствие
ионных компонентов, воды и д р .).
Проявление тех или иных эффектов при взаимодействии электронов с веществом
зависит от энергии электронов. Для усредненного состава продуктов деления, посту
пающих в отходы, средняя энергия излучения (Еср) может колебаться в пределах
0,3-1,0 МэВ, когда превалируют эффекты возбуждения связанных электронов и иони
зация внешних и внутренних электронных оболочек атомов.
В алюмосиликатных минералах со сложными кислородосодержащими анионами
связи между элементами решетки достаточно сильны, объем свободных междоузлий
мал и это определяет небольшую вероятность смещения атомов из нормальных поло
жений. Зафиксированы переносы электрона от аниона к катиону, как главное прояв
ление действия излучения на алюмосиликатных минералах, но возможен и прямой раз
рыв химической связи внутри иона. Таким образом, кроме изменений в микрострук
туре поверхности (при внешнем облучении и глубине воздействия в несколько сотен
ангстрем), возможно изменение энергетических параметров структуры, т.е. переход
к иной структуре с иными, чем исходные, сорбционными параметрами. Кроме того,
в случае захоронения высокоактивных отходов, учитывая длительность и постоянство
облучения, а следовательно, и огромные значения поглощенных доз, в прогнозах сле
дует также учитывать явление отжига радиационных дефектов вследствие возможно
сти его инициирования за счет термического воздействия.
Проведенные нами исследования показывают, что облучение сухих алюмосили
катных систем практически не меняет их суммарных сорбционных емкостей по отно
шению к радиоизотопам даже при значительных величинах накопленных доз ( 1 0 11 рад) .
Облучение систем, содержащих влагу или растворы, приводит к небольшому увеличе
нию сорбционной емкости уже при дозах 4,2- 108рад и выше, что вызвано появлением
дополнительных активных центров при одновременном изменении поверхности сорбента.
Методом ИК-спектроскопии установлено уменьшение относительной интенсивно
сти полос поглощения в области валентных колебаний структурных гидроксильных
групп, связанное либо с частичным растворением октаэдрического кремния, либо с
образованием аморфного кремнезема, либо с процессом дегидроксилизации. При об
лучении как воздушно-сухих, так и влагонасыщенных образцов происходит измене
ние кинетики сорбции. Равновесие при облучении устанавливается быстрее, что свя
зано с изменениями в поверхностном слое. Для высококристалличных глин структу
ра поверхности изменяется с возрастанием кислотности.
Проведено определение влияния ионизирующего излучения на сорбционные
свойства отдельных минералов алюмосиликатных пород: кварца, каолинита, полево-
IAEA-SM-243/113 189
ТАБЛИЦА VI. ИЗМЕНЕНИЕ СОРБЦИОННОЙ ЕМКОСТИ ПОРОД
ПРИ ОБЛУЧЕНИИ (мощность дозы - 480 рад/с)
Обработка образцов Доза(рая)
Сорбционная емкость ,0Sr (мг/100 г)
Облучен в растворе 6,9 - 103 1,86
HNOj - 20 г/л, 2 ,9 -104 2,0
N aN 03 - 150 г/л 1,7 -10* 2,15
4 ,7 -1 0 ' 2,27
4 ,2 -10“ 2,61
8 ,3-10“ 3,06
го шпата и хлорита (60Со, установка MPXy-lOO, без нагревания (20°С) и при нагрева
нии (50°С) , электронный ускоритель прямого действия с энергией электронов 1 МэВ) .
Облучение проведено при мощностях доз источника МРХ 7 -IOO, равных 1,8; 2,5;
3,1Мрад/ч,и480рад/ч—на электронном ускорителе. Идентификация изменений в образ
цах после облучения проводилась методом ИК-спектроскопии и дифрактометрии. Ме
тодом меченых атомов с использованием ^ S r изучали изменение кинетических пара
метров сорбции на минералах.
Полученные результаты свидетельствуют о том, что под действием облучения в
минералах появляются дефекты в поверхностных слоях. Эти дефекты связаны с де
гидратацией и дегидроксилизацией, с разрушением силоксановых и силонольных свя
зей (S i — ОН; Si — О — Si) . Глубина перераспределения связей поверхностного слоя
зерен не превышает нескольких сотен ангстрем, но для каждого минерала несколько
отлично, что определяется индивидуальностью параметров связей между атомами в
структуре решетки [5].
Результатом действия облучения в пределах изученных доз являются изменения
в сорбционных свойствах: ускоряется межфазное распределение нуклидов и увели
чивается прочность их локализации в твердой фазе. В табл. VI приведены данные из
менений сорбционной емкости алюмосиликатных пород при облучении. Указанные
изменения относятся к незначительной части каждого образца пород, главным образом,
к слоям в несколько сотен ангстрем, в целом же породы остаются радиационно-устой
чивыми.
190 СПИЦЫН и БАЛУКОВА
6. ОЦЕНКА ПРОВЕДЕНИЯ ТРАНСУРАНОВЫХ ЭЛЕМЕНТОВ
В ПЛАСТАХ-КОЛЛЕКТОРАХ
В реальных отходах содержание трансурановых элементов меняется от десятых
долей миллиграмма до предельно допустимых концентраций. Формы существования
нуклидов зависят от ’’биографии” отходов, присутствия макрокомпонентов и радио
лиза растворов. Изменения форм существования происходят во времени как в самих
отходах, так и в пластовых условиях и наибольшее влияние имеют радиационно-хими
ческие превращения. Поскольку в пластовых условиях с течением времени неизбеж
ны процессы гидролиза практически всех комплексных соединений, все катионные
формы нуклидов достаточно надежно удерживаются в твердой фазе. Сорбция а-нук-
лидов из кислых растворов проходит с коэффициентами распределения для различных
их форм от 1,1 до 4. Накопление плутония составляет до 60 мг на 1 кг породы, неп
туния — до нескольких мг/кг, а америция — десятки мкг/кг. Десорбируемость фильт
рующим потоком затруднена ввиду того, что параллельно протекают процессы гид
ролиза макрокомпонентов, маскирующих сорбционное накопление а-нуклидов. Вслед
ствие сорбции на минералах и гидролиза макрокомпонентов очистка растворов от них
по фронту движущегося потока практически полная.
ЛИТЕРАТУРА
[ 1] СПИЦЫН, В.И. и др., ’’Основные предпосылки и практика использования глубоких водоносных горизонтов для захоронения жидких радиоактивных отходов” , Nuclear Power and its F u e l C y c le , v. 4 (P roc . In t . C o n f . S a lzb urg , 1 97 7 ) IA E A , V ie n n a (1 9 7 7 ) 4 81 .
( 2] СПИЦЫН, В.И., БАЛУКОВА, В .Д ., Исследования в Области Обезвреживания Жидких,Твердых и Газообразных Радиоактивных Отходов и Дезактивации Загрязненных Поверхностей, т.II (Научн.-техн. конф. СЭВ, Колобжег, 1973) Варшава (1973) 7.
[ 3] СПИЦЫН, В.И., БАЛУКОВА, В.Д., ЕРМАКОВА, Т .А ., Исследования в Области Обезвреживания Жидких, Твердых и Газообразных Радиоактивных Отходов и Дезактивации Загрязненных Поверхностей, т.1 (Научн.-техн. конф. СЭВ, Колобжег, 1973) Варшава (1973) 91.
[4 ] ЕГОРОВ, Е.В., НОВИКОВ, П .Д ., Действие Ионизирующих Излучений на Ионнообменные Процессы, Атомиздат, М., 1965.
[ 5] БАЛУКОВА, В.Д., КОСАРЕВА, И.М., Исследования в Области Обезвреживания Жидких, Твердых и Газообразных.Радиоактивных Отходов и Дезактивации Загрязненных Поверхностей, вып. 3 (Научн.-техн. конф. СЭВ, Москва, 1977) Атомиздат, Москва (1978) 29.
DISCUSSION
J. A ROD: You spoke about the possibilities of radionuclide retention in
geological formations. Could you say something about the risks of radionuclides
thus fixed being displaced by the action of cations contained in groundwater or
the action of macromolecules of the humic or fulvic acid type which are also
present in such water?
IAEA-SM-243/113 191
Valentina BALUKOVA: The description of the mechanism of the sorption
and migration processes given in the paper takes into account the salt composition
of the formation water and wastes. Sorption was studied in the presence of various
organic complexes, including the acids you mentioned. Taking all the factors into
consideration, there is no risk of displacement of the sorbed radionuclides. Puri
fication of the movement front of the liquid phase of the waste is effected below
the maximum permissible concentrations, and the waste movement is regularly
monitored.
C.J.G. BARRAUD: After injection of radioactive solutions into the formation
did you observe a significant increase in the temperature of the formation water?
Valentina BALUKOVA: In the case of intermediate-level waste, the tempera
ture rose by several tens of degrees (from 14 to 50°C).
C.J.G. BARRAUD: In cases where the temperature rose by about fifty degrees,
did you observe mineralogical changes in the solid phase of the disposal formation?
Valentina BALUKOVA: Such temperatures do not lead to changes in the
mineralogy as far as the accumulation of radionuclides by sorption is concerned.
The sorption kinetics are improved.
IAEA-SM-243/66
ALMA - A STUDY OF A REPOSITORY
FOR LOW- AND MEDIUM-LEVEL WASTE
IN A ROCK CAVERN
N. RYDELL, O. DEGERMAN
National Council for Radioactive Waste,
Stockholm
C. THEGERSTRÔM, R. GELIN
Studsvik Energiteknik AB,
Nykôping
M. CEDERSTRÔM
Swedish State Power Board,
Vàllingby,
Sweden
Abstract
ALMA - A STUDY OF A REPOSITORY FOR LOW AND MEDIUM-LEVEL WASTE IN A ROCK CAVERN.
A conceptual study has been made of a repository for medium-level waste in a rock cavern at shallow depth. The purpose of the study has been to verify that this kind of waste can be safely disposed of in such a repository and to analyse different designs with respect to safety and construction costs. The properties of a typical host rock for the repository are described and used to derive the requirements for containment capability of the repository.A design has been made for the total volume of medium level waste for 30 years operation of 10 000 MW(e) generating capacity. The construction costs of this repository have been estimated to add 0,03 mffls/kW-h to the power costs. The design provides for peripheral barriers in the repository which satisfy, usually with wide margins, the requirement that water extracted from a well drilled in the rock above the repository shall contain less than MPC„, of any nuclide from the waste.
1. BACKGROUND
Six nuclear power units are in operation in Sweden at present and six more
are contracted for and at various stages of implementation1. The total generating
capacity for these twelve units is about 10 000 MW(e). They are located at four
coastal sites.
1 The near-term development of this programme will be the subject of an advisory referendum. Subsequent to this referendum the present programme may be confirmed or modified.
193
194 RYDELL et al.
TABLE I. WASTE VOLUMES
Waste type
Total waste, volume
(m3)
Medium-levelwaste,volume(m3)
Medium-levelwaste,total surface (m2)
Concrete cubes 70 000 55 000 275 000Bitumen drums 20 000 16 000 160 000
Ash drums 6 000 2 000 20 000
Compacted solids 24 000 7 000 65 000
Total 120 000 80 000 520 000
Nuclear wastes from the operating plants are currently conditioned and
stored at the sites, with the exception that burnable waste is transported to
Studsvik for incineration and storage of the ashes. The sites were, however,
never intended for final disposal of these wastes nor does the idea of four waste
repositories seem attractive when the total amount of medium-level waste is
well within the capacity of one. It is thus justified to plan for one central
repository for all medium-level waste from our nuclear power programme.
The National Council for Radioactive Waste has undertaken a conceptual
study of a central repository named ALMA. The purpose of the study has been
to verify that medium-level waste can be disposed of without undue risk to
future generations, and to analyse different design alternatives for aspects of
safety and construction costs.
The study is based on the currently scheduled nuclear power programme
and deals therefore with all medium-level waste arising from 30 years operation
of 10 000 MW gross nuclear electric generating capacity.
2. WASTE CHARACTERISTICS
There are no generally accepted definitions of low-, medium-, or high-level
nuclear waste. For the purpose of this report on disposal we define medium-
level waste as such waste which after 100 years still contains significant amounts
of radioactive nuclides, but which has negligible thermal energy release.
The end-products form the waste treatment; at the nuclear power plants
are concrete cubes, 200 ltr steel drums containing bituminized waste or ashes
and compacted solids in steel containers.
IAEA-SM-243/66
TABLE II. RADIOACTIVE NUCLIDE INVENTORY
195
Nuclide Half-life(a)
Activity(a)
H-3 12.3 300
C-14 5 735 2
Co-60 5.3 20 000
Ni-5 9 80 000 50
Ni-63 92 500
Sr-90 28 10 000
1-129 17 000 000 0.5
Cs-135 3 000 000 0.5
Cs-137 30 100 000
Pu-239 24 400 0.5
The total waste volume from 30 years operation of 10 000 MW nuclear
electric power is estimated to be about 120 000 m3. Table I gives the volumes
of the different end products and the total surface of the waste containers.
The overall content of radioactive nuclides in this waste has been estimated
with the conservative assumption that 10~3 of the fuel pins in the cores are
leakers. Table II gives the estimated inventory. Ci-values are calculated from
measured data in our plants for the more abundant nuclides and from literature
data for the less abundant but extremely long-lived nuclides.
Low- and medium-level wastes from reprocessing have not yet been
characterized as to waste forms, volumes and radioactive content. No difficulty
is foreseen in accommodating the volumes, but the content and leachability of
the actinides in different categories of reprocessing waste will have to be weighed
against the containment capability of the barriers around the waste before it
can be decided to what extent such waste can be deposited at shallow depth. It
has nevertheless been assumed in the design of the repository that one third of
the total waste volume will come from reprocessing.
3. GEOLOGIC CHARACTERISTICS OF THE MODEL SITE OF THE
REPOSITORY
Shallow land burial in sand or clay is in many countries considered as
appropriate for low- and medium-level waste. Geologic conditions in Sweden
196 RYDELL et al.
IAEA-SM-243/66 197
CRUSH ZONE
REPOSITORY OUTLINEFRACTURE ZONES
A, В and C-wells
FIG.2. Configuration of repository and wells. Repository is located under a rock cover o f 50 m.
are, however, not suited for this type of disposal of medium-level waste. Loose
sediments more than 15 m deep are rarely found in our country except in river
valleys and in those areas which were below sea level after the last glaciation.
These areas are extensively used for farming and housing.
Crystalline rock of good quality is, however, available at or near the surface
over most of the country and can be excavated at reasonable cost. Crystalline
rock has therefore been selected as a geologic environment in our study of a waste
repository.
We have taken as a model of a good repository site, a rock body 0.5—1 km2
in size with a generally flat surface topography and bounded on all sides by
severely fractured zones. Such rock bodies are a normal feature of the Swedish
bedrock (Fig. 1). They are usùally traversed by a few fracture zones but other
wise of excellent quality for excavation work. The fracture zones are of little
detriment since they can be sealed by grouting of the cracks.
Radioactive nuclides which leak from the repository will be transported by
groundwater in rock fractures to aquifers downstream in the hydraulic gradient
field and ultimately to the sea. Of importance for dose calculations may be
transport times versus decay of radioactivity, retardation due to sorption on
fracture surfaces and in the rock mass, dilution of the nuclides in gradually larger
water volumes and man’s utilization of the contaminated groundwater.
Dilution of this groundwater in a stream or a lake will greatly reduce the
concentration of the contamination. Water in wells drilled in the bedrock along
the groundwater path downstream from the repository will be more contaminated.
Restrictions may be put on such well drilling in the vicinity of the repository
for a limited time but not for ever.
We have therefore in the analyses of the safety of the waste disposal concept
conservatively assumed that a well is drilled near the repository. Three qualitatively
different locations of such a well have been selected (Fig.2).
198 RYDELL et al.
Wells A or B are of limited capacity for natural reasons and are assumed
to serve only a few households. Well С has a large capacity and is assumed to
have a correspondingly large utilization.
The hydrological flow model for this rock body should take into account
that the groundwater flow is limited to cracks and fractures and that the
distribution of these is often anisotropic and site specific. A generalized model
for the analytic treatment of these conditions is, however, not available. We
have therefore calculated flow patterns as if the hydraulic conductivity of the
rock were isotropic and homogeneous, and then adjusted the results in cases
where this simplification obviously is misleading.
The relevant parameters of the host rock are hydraulic conductivity and
rock porosity and of the well are depth, capacity and consumption.. Our
assumptions on rock parameters are based on field data obtained by the Swedish
Geologic Survey. They are listed in Table III.
Calculation with potential theory for groundwater flow in homogeneous,
isotropic, porous rock shows that the fully developed cones of depression of
well A and well С will influence the groundwater flow so that it is drawn to
the well over a distance which goes beyond the outline of the repository. The
groundwater flow will therefore theoretically collect all nuclides which leak
from the repository. This is not unrealistic in the case of well C, since the
crush zones will anyhow drain all water which inflitrates the ground over the
repository and since we have not assumed any pronounced topographical
gradient.
It is unrealistic in the case of well A, though. A rock body of this size
is bound to have discontinuities in the form of fracture zones as indicated in
Fig. 2. The fracture zones have considerably higher conductivities than the
rock near the well as indicated in Table III. The flow pattern will be
concentrated in such fracture zones. The rock volume on the other side of
the fracture zone will be much less influenced by the well than in the homo
geneous model. In regard to the dimensions of the repository, 300 X 25 X 30 m,
and the frequency of more or less developed fracture zones in crystalline rock
it seems reasonable to assume that the dominating influence of well A on the
groundwater flow will only extend over 10% of the length of the repository.
The transport times of the ground water from the periphery of the
repository to the wells A and С have also been evaluated. They are in both
well alternatives less than one month.
4. PERMISSIBLE LEAK RATES FROM THE REPOSITORY
These data on the properties of the host rock make it possible to
evaluate permissible leak rates of different radioactive nuclides from the repository.
IAEA-SM-243/66
TABLE III. PARAMETERS OF ROCK AND WELLS
199
Cond.(m/s)
Porosity Depth(m)
Capacity(m3/s)
Consumption(m3/s)
Well A in
undisturbed rock 1(T8 io -4 50 2 X 10~5 2 X 10"5
Well В in
fracture zone 1СГ6 5 X 10' 4 25 5 X 10"4 2 X 10' 5
Well С in
crush zone 10~5 10 '3 50 10'2 10 '2
This requires, however, a definition of acceptable contamination of the well
water. We have used the MPCw-values recommended by ICRP for water which
may be used for drinking.
The activity concentration c¡ Ci/m3 in the well water may in the first
approximation be expressed as
ci= S i 'L i ' f r ' ag /Q (1)
where S¡ = source strength in curies of nuclide i, L¡ = leak rate (s-1) of nuclide i
from the repository, fr = fraction of repository periphery from which leakage is
collected by ground water flowing to the well, ag = attenuation of the nuclide i
concentration during its transport to the well due to radioactive decay and
sorption, Q = water inflow to equal water yield from well (m3/s).
Permissible values of L¿ are obtained if c¡ is set equal to MPCW¡, S¡ is taken
from Table II. fr is 0.1 for well A and 1.0 for well C. ag is set equal to 1.0
because of the comparatively short transport times and because we are only
considering nuclides with half-lives very much in excess of one month. Q is
taken from Table III.
Values of L¿ are listed in Table IV. We have in addition given values on Lwi
permissible leak rates from individual waste containers per m2 surface area and
year. The Lw¡ values refer only to the particular waste inventory which is listed
in Table I. They indicate the requirements that would be put on waste contain
ment quality if no extra leakage barrier were provided between the waste con
tainers and the periphery of the repository.
TABLE IV. PERMISSIBLE LEAK RATES
200 RYDELL et al.'
NuclideRepositoryinventory(a)
MPC*,
(a/m3)
Q/f
(m3/s)
Li
(s'1)
Lwi
(a/m2 -a)
H-3 300 3 X 10"3 10~2 10~7 6 X 10' 6
C-14 2 8 X 10~4 2 X 10-4 8 X 10~8 5 X 10“6
Co-60 20 000 3 X 10~5 10~2 1.5 X 10~u 9 X 10~10
Ni-5 9 50 2 X 10~4 2 X 10~4 8 X 10~10 5 X 10' 8
Ni-63 25 3 X 10~s 2 X 10~4 2.4 X 10' 10 1.5 X 10' 8
Sr-90 1 000 4 X 10' 7 2 X 10' 4 8 X 10' 14 5 X 10' 12
1-129 0.5 4 X 10' 7 2 X 10~4 1.6 X 10~10 10~8
Gs-135 0.5 1 X 10‘ 4 2 X 10' 4 4 X 10' 8 2.5 X 10~6
Cs-137 10 000 2 X 10~5 2 X 10~4 4 X 10-13 2.5 X 10"n
Pu-239 0.5 5 X 10' 6 2 X 10“4 2 X 10~9 10~7
We have further assumed that no well drilling will be permitted over the
repository during the first 100 years, but that no restrictions are imposed on
well drilling in the adjacent fracture zones. L¡ and Lwi will depend on which
well is considered. Table IV lists the most restrictive case.
The values of L¡ and LW1 must not be taken as characteristic of medium-
level-waste repositories in shallow rock caverns. They have been derived in a
conceptual study in order to illustrate the type of analysis that can be made
and in order to indicate to what extent, if any, further leakage barriers should
be included in the design of the repository.
5. REPOSITORY DESIGN
The magnitude of the permissible leak rates in Table IV indicates the
needed degree of containment of the waste. This will either have to be provided
by the inherent containment quality of each waste package or by a common,
peripheral barrier. The advantage of a common barrier is obviously that it
reduces the quality requirements on the packages and the need to verify that
the requirements are met on each and every package. The cost saving can be
substantial when the number of packages is more than 100 000, as in our case.
IAEA-SM-243/66 2 0 1
The repository design should be such that:
no supervision is needed once the repository is sealed;
the protective function of any leakage barrier which is accounted for in
the safety analysis can be verified during the construction and predicted
for the required lifetime;
no interaction takes place between the rock, the groundwater, the barriers
and the material in the waste packages which could conceivably cause
significant deterioration in the function of the barriers;
it permits expedient handling and transport of the many waste packages so
that radiation exposure of the personnel is minimized;
the net impact of the disposal on the total costs of waste management is
minimized.
Three designs of the storage caverns have been evaluated. They are:i
(1) several short tunnels (Fig. 3);
(2) one large cavern (Fig. 4);
(3) several deep vertical shafts with cylindrical cross-sections (Fig. 5).
The leakage barrier consists in all three cases of a layer of sand-bentonite
mixture. This layer is backed up in alternatives (2) and (3) by a concrete wall.
The sand-bentonite layer is ductile and will have a low hydraulic conductivity,
about 10-12 m/s. This prevents any flow of groundwater through the repository
and leaves diffusion as the only transport mechanism for radioactive nuclides.
Concrete is an excellent diffusion barrier and clay an excellent sorbent for ground
water impurities. The tendency of large concrete constructions to crack is offset
by the self-sealing property of clay.
The bentonite fraction of the mixture should be 10 to 15%. The lower
fraction has a higher load-carrying capability and is therefore used in the bottom
of the repository where a high compaction pressure is maintained by the
gravitational load of the waste pile. The higher bentonite fraction gives a larger
swelling and self-sealing ability to the mixture when and where it is penetrated
by groundwater. It is used in the backfilling of the top of the cavern over the
waste pile. The filling at the sides of the cavern holds an intermediate bentonite
fraction.
The tunnel alternative is flexible since different tunnels can be used for
different kinds of waste. It is, however, difficult to take advantage of this
flexibility in an actual case.
The disadvantages of the tunnel alternative is that the transport of the
waste is made with trucks, which is a cumbrous operation during which it is
difficult to protect the truck driver against radiation. The excavated volume
2 0 2 RYDELL et al. '
SECTION A-1:200-
FIG.3. Tunnel alternative.
is not utilized very efficiently in the tunnel alternative because of the narrow
dimensions of the tunnels and the need to reserve peripheral space for the
clay barrier. It is also difficult to backfill the tunnels properly when no
concrete wall around the waste provides radiation shielding during the back
filling operation.
The large cavern and the vertical shafts give better volumetric utilization of
the excavation and are well suited to handling of the waste with remotely
operated traverse cranes. The concrete walls serve primarily as support for the
traverse crane but are useful in several other respects.
TRAVERSE CRANE
IAEA-SM-243/66 203
-57 m
-61,7 m
-8 0 m
1,5 19,6
24,0
A ll d im en s io n s in m etres .
. B l. 1,5
FIG.4. Cavern alternative.
They stabilize the waste pile and thus permit piling of the waste to con
siderable heights. The final backfilling of the side and top spaces is facilitated
because of the radiation shielding provided by the walls and because the walls
support the compaction pressure which should be applied to the clay-mixture
during the backfilling.
The dimensions of the sand-clay barrier and of the concrete walls have been
determined by construction requirements rather than by the need for adequate
containment. The concrete walls are 700 to 800 mm thick. This provides
sufficient strength to the construction and ample radiation shielding for any
operations which are performed in the space outside the walls. The sand-clay
barrier is about 1.5 m thick. This provides sufficient space for the equipment
with which the filling is compacted.
The alternatives (2) and (3) have been combined in our design study as
shown in Fig. 6, in order to evaluate excavation techniques, transport of broken
rock material and of waste, and provide flexibility in the design. These alterna
tives must be studied further before their technical merits can be evaluated fully.
The cost of the repository of Fig. 6 has been estimated at 60 million US $.
This corresponds to 500 $/m3 waste and to 0.032 mills/kW-h.
The design in Fig. 6 is not the most cost-effective, since it includes two
different types of storage caverns. The cost per m3 waste volume is estimated to
be higher for the vertical shaft than for the horizontal cavern.
204 RYDELL et al.
TRAVERSE CRANE OR GANTRY CRANE
1,5
-1 1 7 4.2_sfe____All.dim cm ions in metres.
FIG. 5. Vertical shaft alternative.
The sand-clay barrier is a safety measure which is added to the inherent
safety of an underground location in good rock. Our analysis indicates that it
is justified, but it might be avoided if the repository is located at greater depth.
The cost of the sand-bentonite barrier is estimated at 10 million dollars. The
cost of going from 50 m depth as in our concept to 300 m depth is estimated
to be considerably higher, 20 million dollars.
6. CONTAINMENT EFFICIENCY OF REPOSITORY
The basic function of the peripheral barriers in the repository is to prevent
any circulation of groundwater through the repository. Radioactive nuclides
can then only be transported by diffusion. The slow propagation of nuclides
STORAGE FOR DRUMS
FIG. 6. Provisional design o f repository.
206 RYDELL.et al.
TABLE V. DIFFUSION COEFFICIENTS, m2/a
Nuclide Concrete Clay
H-3 7 X 10'4 7 X 1СГ2
С 3 X 1(T4 3 X 10"2Fe, Ni, Со 4 X 10'9 8 X 10'6Sr 7 X 1(T6 7 X 10~5I 6X 10~4 6 X 10~2Cs 3 X 10'4 2 X 10'4Pu 4 X 10'9 3 X 1(T6
by diffusion will be further retarded by sorption in the concrete and the clay.
We have made preliminary calculations of the diffusion of significant nuclides in
the waste to groundwater in fractures in the surrounding rock. The diffusion
coefficients used in the calculations are given in Table V.
It has so far not been possible on available computer models for diffusion
calculations to make a satisfactory representation of the transition from the
clay to the cracks in the rock, which typically have widths of 0.1-0.2 mm. The
rock has instead been represented in the model as homogeneous with a porosity
of 10”4 or heterogeneous with 2 mm wide cracks and a porosity of 2 X 10-3.
The homogeneous model gives, as expected, a higher and earlier peak in the
leakage rate than the heterogeneous model. It is thus conservative in this
respect. It should be observed, though, that an overestimation of the initial
leakage rate gives an underestimation of the inventory at later times when the
efficiency of the barriers may have deteriorated.
The following conclusions can be made at present. Virtually no 137Cs, 90Sr
or radioactive Fe or Co will ever leave the repository. The calculations give the
same result for Pu and 135Cs but further investigations should be made on long
term chemical conditions in the repository before it can be concluded that the
quoted diffusion coefficients are valid. 3H and 14C will leak out in insignificant
amounts.
129I is the only nuclide which leaks out in amounts of any significance. The
leak rate from the repository peaks at about 3 X 10"13 Ci/s or 10 microcuries
per year after 1000 years. This gives a margin of two to three orders of
magnitude to the permissible leak rate in Table IV.
Such a repository thus meets the requirements put down in Table IV with
a wide safety margin.
IAEA-SM-243/66 207
These diffusion calculations should nevertheless be continued in particular
with a view towards possible long-term changes in the properties of the barriers.
The impact of a saturation of the clay-sand mixture with Ca from the concrete
is one concern. The chemistry inside the repository is another.
7. CONCLUSIONS
Our studies of a medium-level-waste repository in a rock cavern at shallow
depth indicate at the present stage that the costs of this type of disposal,
although considerable as a lump sum, make an insignificant contribution to
the cost per kW -h of nuclear electric power. The provision of a leakage barrier
between the waste and the rock walls reduces any leakages of radioactive
nuclides to insignificant amounts as long as the initial conditions in the repository
remain. Further studies of this concept should be directed towards possible
long-term changes in these conditions.
DISCUSSION
K. KÜHN : In your conceptual design did you consider any accident scenarios
other than well-drilling, taking into account the fact that the repository is covered by .
50 m of rock and that there is also some Pu and 129I in the water?
N. RYDELL: No, we considered well-drilling to be the greatest hazard. A
new glaciation would certainly disrupt the integrity of the barriers but any
leakage of activity would be strongly diluted before it could reach man, and after
the glaciation there would be very little activity left in the repository.
H. KRAUSE: Did you investigate the density which the sand-bentonite
mixture must have in order to prévent any water infiltration into the waste?
How will you ensure that this density is attained during plant operation, especially
between the drums?
N. RYDELL: Yes, we have investigated the necessary densities. In paper
SM-243/22 this question is discussed in some detail. We do not intend to use
the sand-bentonite mixture between the drums. There we plan to inject concrete
to ensure mechanical stabilization of the waste pile. This injection will be made
layer by layer.
H. KRAUSE: Do you think that there is any risk of segregation of sand and
bentonite during introduction of the mixture?
N. RYDELL: We do not anticipate any risk of particle segregation. Again
I would refer you to paper SM-243/22 and to the references cited there.
208 RYDELL et al.
W.R. BURTON: One risk you have considered is that of obstruction of
drinking water. Have you looked into the advantage of siting the repository in
rock under the sea?
N. RYDELL: We decided to study a type of repository site which was
widely available. A repository in rock under the sea is a special site which would
have the advantage that there is no likelihood of future interference by man.
There are, however, two associated problems. First, the diffusion barrier would
be wetted by saline water, which is a disadvantage. Second, since our landmass is
still rising after the last glaciation and since we do not know how much more it
will rise, “under the sea” is not a well-defined criterion.
F. GERA: Is it conceivable that the effectiveness of the geochemical
barrier could be improved by surrounding the waste with a layer made of blocks
of highly compacted bentonite?
N. RYDELL: Blocks of this kind in sufficient quantity to form a barrier
around a repository of our size would be very expensive. I believe such a solution
is conceivable but it seems unnecessary since our bentonite-sand mixture is
sufficiently effective.
K. TIETZE: Have you carried out, or are you planning, pumping tests and
tracer experiments in the special region? I ask this question because for water
extracted from a well drilled in the rock above the repository, you specify a
pollution of less than the maximum permissible concentration in water of any
nuclide from the waste.
N. RYDELL: Extensive hydrological tests, including pumping tests and
some tracer tests, are being conducted at our Finnsjô test site and extensive
tracer tests at Studsvik. The results of these tests will be used to validate our
models. No special region has so far been selected for the repository nor has a
site confirmation programme been decided upon.
IAEA-SM-243/156
DISPOSAL OF LOW- AND INTERMEDIATE-LEVEL
WASTE IN CZECHOSLOVAKIA
Z. DLOUHŸ
Nuclear Research Institute,
Rêz
J. KORTUS
Chemoprojekt,
Prague
E. MALÁSEK
Czechoslovak Atomic Energy Commission,
Prague
J. MAREK
Czech Power Works,
Prague
M. SELIGA
Slovak Power Works,
Bratislava
Abstract
DISPOSAL OF LOW- AND INTERMEDIATE-LEVEL WASTE IN CZECHOSLOVAKIA.The Czechoslovak approach to, and Czech practice in, the storage of radioactive wastes
from research institutes, radioisotope users and nuclear power plant operation are presented.The specific territorial conditions do not permit the dilution and dispersion of larger quantities of wastes. Initially particular attention was given to the concentration of wastes, to reduction of their volume, to transformation into suitable form and to the selection of convenient areas for final disposal. Very good experience has been had with repositories in dry abandoned limestone mines used only for storage of wastes produced by isotope users and research institutes. The available storage capacity of these mines is relatively small. The selection of repositories for wastes from nuclear power plants has involved screening 51 available sites. By means of exclusion criteria the two most convenient sites were selected. Design studies of regional repositories on the sites have been elaborated.
1. INTRODUCTION
Czechoslovakia is a small continental country and because of its unfavourable
hydrogeological conditions and high population density, the dispersion or dilution
capacity of the environment is rather small. The emphasis is therefore on the
209
treatment and concentration of radioactive wastes into the smallest possible volume
and most suitable form for permanent storage in special areas.
In Czechoslovak conditions the principle of centralized disposal of radio
active wastes has been applied from the very beginning.
The first experimental disposal site for wastes arising in production and
utilization of radioisotopes and in research institutes was put into operation in
1959 and was closed in 1963. The central disposal site for the same type of wastes
started operation in 1964 and is still in use. Because of the relatively small volume
of wastes limestone quarries were selected in both cases.
Operation of nuclear power plants is connected with significantly larger
volumes of radioactive wastes and therefore construction of two regional disposal
sites of surface type is under way. After three years of investigation two sites
were finally chosen and the final geological survey of these sites is nearly complete.
Operation of the first regional waste disposal site for radioactive wastes arising in
nuclear power plants will start in 1983.
210 DLOUHY et al. ,
2. OPERATIONAL EXPERIENCE IN DISPOSAL OF WASTES
PRODUCED IN RESEARCH AND UTILIZATION OF RADIOISOTOPES
In selecting a central waste disposal site about 20 years ago several possibilities
were considered: disused mines, different underground structures or surface sites.
With regard to the amount and the character of radioactive wastes from research
institutes, utilization of radioisotopes and the economics of waste transportation,
a system of horizontal galleries in a disused limestone quarry was selected, located
about 30 km from Prague.
In the selection process a large number of old mines were found, but it was
very difficult to find a suitable site for waste disposal. Nearly all the mines are in
contact with the groundwater and the water has to be pumped out during the
mining period. Therefore the costs of maintenance and mine operation together
with the problem of adherence to strict radiological protection rules excluded the
utilization of old mines. Only a few sites complied with the fixed criteria and
two limestone quarries were selected.
An experimental disposal site was put into operation in 1959. Inca ted in
galleries of a disused limestone auarrv situated approximately 100 m from a small
river and 5 m above the groundwater level. The solid wastes were stored in 50-ltr
zinc-coated metal containers; liquid concentrates were mixed with cement in 100-ltr
steel drums coated with bitumen. The disposal of wastes ceased in 1963 and during
the next ten years the properties of the wastes and containers were observed.
The area chosen for waste storage was part of the system of horizontal
galleries without any fissures. It consisted of an access drift (30 m long, 2 X 2 m)
IAEA-SM-243/156 211
and of a cave 5 X 8 X 3 m. The total storage capacity was 120 m3 and the invest
ment costs were about 55 000 Czech crowns. During the operation and checking
period there was no infiltration of water and the only source of water was the
humid air from the ventilation system. The site was not used for disposal of
wastes containing natural radionuclides and no contamination of air was detected
in the disposal area or in the surroundings.
It was planned originally to transfer the wastes to a final disposal site. Taking
into account the very low content of radionuclides in wastes, negligible corrosion
of containers and the geological properties of the site it was decided to close the
site and to leave the wastes in place. The site is checked once a year and no
irregularities have been discovered.
Experience gained at the experimental disposal site was employed in the
construction of the central disposal site. This site was also constructed in a lime
stone formation of the same type.
In 1960 the project and the geological and hydrological surveys were carried
out. The facility was built in the years 1961-1963 and was nut into operation.at
the beginning of 1964.
Permanent storage was provided in a large disused underground limestone
quarry, comprising a complex system of underground corridors with an entrance
above ground. The quarry is in the area of the Bohemian chalk system and the
geological survey showed the following results: the upper layer is a clayey silt
about 2 m thick, then there is a 75 m thick marl laver, a further 5 m of cretaceous
marly limestone and finally marly sandstone. The storage spaces are in exhausted
limestone drifts. The impermeability of the layers above and below the drifts
was ascertained. The state of surface waters is satisfactory; the seepage in under
ground cavities is negligible. The level of groundwater is about 45 m below the
drifts in the sandstone layer and can be checked in the two deepest bore holes.
The site is situated in the basin of the River Labe, which is 1.7 km away.
The total cost of the construction of the facility amounted to 6.15 million
Czech crowns. The underground space of the waste storage facility forms a system
of halls and galleries used for the transportation and storage of containers with
radioactive wastes. The approximate area of the used underground space is
4000 m2 and the volume 10 000 m3, with the possibility of further expansion.
The space is aired by means of a ventilator and an air shaft. Channels and pipes
drain the waters through control pits and a collecting pit and water with a permissible
level of activity is then passed to a river. Solid wastes are stored either in 50-ltr
zinc-coated metal containers or in 100-ltr steel drums; concentrates are incorporated
into bitumen and stored in 100-ltr steel drums coated with bitumen. During 15 years
of operation no contamination of air or seepage water has been detected.
Experience with the storage of radioactive wastes in limestone cavities is very
good, but because of the limited storage area such sites cannot be used for disposal
of wastes from the operation of nuclear power plants. For this purpose the surface
type of disposal site was selected.
2 1 2 DLOUHY et al.
3. THE CZECHOSLOVAK APPROACH TO THE DISPOSAL OF WASTES
FROM NUCLEAR POWER PLANT OPERATIONS
The Czechoslovak nuclear programme is based on the Novoveronesh-type of
nuclear power plant (PWR). The most important problem is to assure safe and
long-term isolation of radioactive wastes from the environment. Since the repro
cessing of spent fuel from PWRs on Czechoslovak territory in the near future is
not under consideration, the main necessity is to resolve the problem of the
management of wastes from nuclear power plants. Gaseous wastes are treated in
the conventional system of decay tanks, iodine filters and absolute filters and only
very limited radioactivity is dissipated into the atmosphere. The discharge of
radioactivity in gaseous effluents is considerably lower than the Czechoslovak
limits. Most of the liquid wastes are concentrated and stored in special stainless
steel tanks, located in concrete cells. This solution is not satisfactory from the
point of view of economics and long-term safety. For this reason, solidification
of liquid concentrates and spent ion-exchange resin into ln w 1р я г Ь я Ы р, m a ss has
been accepted. The development of suitable equipment produced in Czechoslovakia
is under way. The basic part of the bituminization unit is the film evaporator.
The concentrates are mixed with aqueous bitumen emulsion and after evaporation
the homogeneous mass is discharged into the steel drums. Only a controlled small
quantity of radionuclides is discharged in the form of liquid effluents. The volume
reduction of solid wastes by means of incineration and bailing and incorporation
of the ashes produced into bitumen has been proposed.
Wastes will be transported in specially constructed containers by rail truck
or road truck. It is proposed to built two regional repositories in Czechoslovakia,
one for Bohemia and Moravia and another one for Slovakia.
4. SITING OF REGIONAL REPOSITORIES FOR THE
RADIOACTIVE WASTE FROM NUCLEAR POWER PLANTS
The principle objective in siting radioactive waste repositories from the point
of view of nuclear safety is the protection of the public from undesirable releases
of radioactive material to the environment. Therefore, the process of siting of
these repositories on Czechoslovak territory has been directed to the selection
of a site where the potential release of radioactive material is as low as reasonably
achievable, taking into account social and economic considerations.
The standard method for assessing radiological protection in the design and
construction of nuclear facilities laid down by both the Czech and Slovak Ministries
of Public Health by law in 1977 specifies the dose limits and dose limit equivalents
for individuals and small groups of the population living in the vicinity of these
facilities, these limits being identical to those shown in ICRP recommendations.
IAEA-SM-243/156 213
For large groups of the population and for the whole population the guiding value
of 40 mSv/1 MW(e) per year for the collective dose equivalent shall not be exceeded.
This guiding value has been derived under the assumption that in the year
2000 in Czechoslovakia nuclear power plants with a total power of 35 000 MW(e)
will have been installed. The mean dose from non-nuclear sources, in Czechoslovakia
represents a value of about 2 mSv/a. The contribution from nuclear energy should
not exceed 0.1 mSv/a. Taking into account the fact that 99% of this value is due /о
to atmospheric discharges of radioactive material into the environment and 1%
to liquid discharges into the hydrosphere, this limit of 1% for potential releases
from radioactive waste repositories seems to be reasonably low.
The selection of suitable sites for shallow land burial of radioactive waste
consisted of three stages. In the first stage a screening procedure was applied in
order to select from larger regions all acceptable sites excluding unsuitable or
less suitable areas by use of so-called exclusion criteria.
Exclusion criteria used in the screening process were of two types. Criteria
of the first degree qualify any property of a region which unambiguously excludes
the possibility of its use. Here should be mentioned regions of potential landslides,
surface faulting, slope instability, subsidence, earthquakes, soil liquefaction on the
one hand, and protection zones of mineral or drinking water on the other hand.
Criteria of the second degree enable the selection of a conditionally acceptable site
where a repository can be built under the assumption that there is an acceptably
economic technical or engineering solution, or that a legal exception can be made
and that the land can be rededicated to the repository. Here regions with high
water tables are included as well as regions with potential for flooding, state
reservations, regions with important agricultural land etc.
In the second stage, i.e. during the site selection, another set of criteria was
applied, these being characterized as evaluation criteria. The objective of this
stage was to decide in favour of individual sites in the light of some of their specific
properties, namely considering these aspects:
( 1 ) Geography
Less acceptable sites are those in the vicinity of large population centres or
sites at higher levels (in mountainous regions), with uneven terrain and
higher precipitation.
(2) Hydrology
Sites in the vicinity of surface waters such as rivers, lakes, pounds etc.,
and sites in inundation areas are considered inconvenient.
(3) Geology
Geologic structures considered acceptable are sites with low permeability
or an impermeable bed, with good draining possibilities, and sites with high
214 DLOUHŸ et al.
sorption and retention properties of soils, above all to long-lived radionuclides
such as 90Sr and 137Cs.
(4) Geohydrology
Preferred are sites with a slow movement of groundwater and with an
advantageous direction of the groundwater flow related to sources of
drinking water.
(5) Meteorology
This aspect is aimed to favour sites in less humid regions where there is
low potential for severe precipitation, which could lead to a flooding of the
repository.
(6) Economics
Economic evaluation includes transportation costs, investment costs related
to the selected repository, etc.
In the third stage of a site evaluation a simple hydrogeological model was used.
In order to implement such a model the following basic inputs are necessary:
(1) the determination of a source (i.e. of the amount of radioactive material
released from the repository);
(2) the determination of the design basis accident at the repository.
Very conservative and unfavourable assumptions were used for the calculation.
As a design basis accident such severe precipitation was chosen as would lead to
a complete flooding of all storage areas, in combination with a failure of the
drainage system leading to a long-term contact of wastes with water for a period
of about 100 hours. Moreover, assuming 30 years’ operation of the repository
10% of steel drums were assumed to be disturbed together with a rupture of the
concrete insulation and of the asphalt coating.
Similarly, the most unfavourable input data concerning sorption and migration
properties of the soil, of the velocity of the groundwater flow, etc. was used.
The final evaluation consisted of the determination of
the total amount of radioactive material released into the ground;
the time necessary to transfer radioactive material to the nearest point of
water use;
the concentration of radioactive material at the most important points of
water use;
the radiological impact due to water and land use in the nearby environment.
In spite of the use of such most unfavourable data characterizing an accident
condition, though with a very low probability, it has been found in selecting various
sites that the radiological impact ranges between 0.01 and 0.05 mSv/a per head, this
value being low compared to the guiding value laid down in Czechoslovak legislation.
IAEA-SM-243/156 215
5.1. Wastes for a permanent repository
For permanent disposal, only solid and solidified wastes enclosed in steel
drums are considered. Standard drums having a diameter of 600 mm, a height
of 859 mm and a useful volume of 200 ltr will be used.
The solidification of liquid wastes and ion-exchange resins will be carried
out by means of bituminization approximately after one year of storage. Solid
wastes are assumed to be burnt and the ashes to be incorporated into the bitumen,
moldable wastes will be pressed directly in steel drums, the remaining wastes will
be packed into steel drums, possibly after having been cut into small pieces.
5.2. Transportation of radioactive wastes to the repository
In designing transport containers for drums the recommendations of the
IAEA were considered. A decisive Criterion was the exposure rate on the surface
of the containers i.e. under 14.33 nA/kg (200 mR/h). The size of the containers
was limited by their total weight in relation to the loading capacity of the postulated
means of transporting and handling and by the carrying capacity of the handling
cranes being used. For structural reasons, cast iron was chosen as a shielding material.
For the transportation of the drums with higher active ion-exchangers incorporated
into bitumen, a cast iron container of cylindrical type weighing 8 t, with a with
drawable cover and shielding 160 mm thick, was proposed. In the container, one
drum only should be transported. Medium-active wastes - evaporator concen
trates — also incorporated into bitumen, are transported in a container with
a transport capacity of four drums. The container is of a cylindrical type having a
shielding of 40 mm of steel, and weighing 5 t. Containers for low-active ion-
exchangers and solid wastes are of a prismatic type with space inside for loading
eight drums. The weight of the container is 2 t; shielding is not required.
Concerning railway transportation, four containers with higher-active wastes
or six containers with medium-active wastes or six containers with low-active wastes
are assumed to be placed in one truck.
Concerning road transportation, two containers with higher-active wastes or
three containers with medium-active wastes or four containers with low-active
wastes can be placed on the semi-trailer. The driver’s cabin is shielded by a steel
plate so that the rate input does not exceed 20 /nSv/h (2 mrem/h).
5.3. Repository
The repository itself is designed with two series of reinforced concrete double pits,
each double pit having the axis dimensions of length 2 X 18 m, width 6 m and
5. DESIGN OF A REGIONAL REPOSITORY
216 DLOUHY et al.
height 5.3 m. The bottom and side walls of the pits are insulated by bitumen
from soaking by water. On the side walls of the pits there are fixed crane tracks
on which travel a stripping crane and a light sliding roof for temporary covering
of the pit at times between filling.
The pits are covered with standard panels having a width of 6 m. Manipula
tion of the panels is carried out by means of a crane. The pits having been filled
up, the crevice between the panels is stuffed with concrete, a waterproof insulation
of the surface is provided and on top of this a protective cement coat is added.
Finally, the pit is covered with 60 cm of soil. Between the series of pits there are
concrete communication pathways.
5.4. Manipulation of the repository wastes
Transhipment of the containers from the transport vehicles is carried out
by means of a full-portal crane with an overhanging crane way. The containers
are deposited at the point of manipulation or directly on the platform vehicles
by means of which the containers are transported to the stripping crane, moving
along its own track.
On the platform vehicle the closure of the mechanical system assuring the
mechanical integrity of the containers during transportation, is manually desecured
and the power supply to drive the mechanisms that unlocks the safety closures
of the containers is connected. The closure being released, it is withdrawn by a
full portal crane and the open container is transferred to the stripping crane of
the repository. If a distant pit is being filled, a special pallet will be placed on
the vehicle and the stripping crane will transport the whole container as far as
the pit that is going to be filled.
The drums are deposited in the storage pit above each other in six layers.
The stripping crane is equipped with tongs with a joint for clamping the drums
and with a telescopic mast for the lift. This enables storage of the drums in an
exactly defined place and manipulation even of toppled drums.
For transportation of the containers on the pallets and for manipulation of
the panels, the crane is equipped with a second trolley having a carrying capacity
of 12.5 t.
All manipulation of radioactive wastes in the repository is remotely controlled
from the driver’s cabin, which is shielded from the radiation. The cabin is equipped
with a lead glass window and with an industrial television set.
5.5. Drainage of rain and soak water from the repository
Panels, covering the storage pits are put sloping towards the dividing walls
of a series of pits, parallel to their longitudinal axis. Rainwater from the pit
surface is led to a lower dividing wall and from here by means of a flume in the
IAEA-SM-243/156 217
upper surface of this wall and through effluent pipes into the surface drain, with
free discharge outside the repository.
The pits are based on a gravel sand coat of 30 cm, placed on a layer made
from broken stone and asphalt, under which there is a layer of natural clay
at least 50 cm thick. The groundwater level is at least 2 m under the lowest part
of the insulation layer. The impermeable base slopes toward the collecting pipeline,
whereby the subsoil of each pair of pits is sloped separately. The side spaces round
the pits are also filled with gravel sand so that the soak water is led by this drainage
system into the collecting pipeline through the control shafts into the concrete
catch basins. A check of the quality of the soak water is carried out by taking
water samples from the shafts; a final check is carried out from the concrete
collecting tanks.
5.6. Escape of radioactivity and protection of the environment
The one probable way through which radioactive matter could escape from
the storage pits into the outside environment is via transportation through the
soil and via underground water into wells and streams. Irrigation of agricultural
soil could lead to its contamination and hence to the contamination of agricultural
products. In order to prevent the escape of radioactivity and contact of the wastes
with water there exists a containment formed by the following barriers:
The first barrier is a container consisting of a steel drum with a wall thickness of
1 mm, protected inside and outside by anti-corrosion coats. Though this
container will be damaged during long-term storage, it forms a necessary barrier
in the initial period when the activity of the wastes is greatest and when any
permeability of the pits should be evident if it should ever occur.
The second barrier is a fixing medium — bitumen — with a leaching rate of the
order 10~5 to 10~6 g/cm2 per day whereby the leaching of radioactive material
is slight in comparison with that of non-active salts.
The third barrier consists of ferroconcrete pits with insulation against the pene
tration of rainwater.
The fourth barrier is the drainage system around the pits by which the soak water
is led through shafts into the concrete collecting tanks where a final check is made.
The fifth barrier is a relatively or totally impermeable base layer with high
retentive abilities preventing the release of radioactivity into the surroundings of
the repository.
2 1 8 DLOUHŸ et al.
A structural solution for the repository can assure safe storage of radioactive
wastes for the whole storage period necessary. The probability of dispersion of
radioactive matter which would expressly influence the radiation load of the popula
tion living in the vicinity of the repository, is under normal repository conditions
nil and in extraordinary situations imperceptible.
DISCUSSION
K.H.O. SAARI: In Section 1 of the paper you indicate that high pumping
costs precluded the use of any abandoned or existing mine as a repository. Were
there any other reasons, and how accurately did you investigate the different
mines?
J. KORTUS: Yes, there were other reasons, namely high investment and
maintenance costs, difficulty of handling the containers in narrow mine corridors
and uncertainty about re-use of the mine region. Preliminary studies have been
made to evaluate the economics and safety of uranium mines and mines of other
types.
G. STOTT : What is the specific activity of the concentrated liquid waste
in the stainless steel tanks?
J. KORTUS: The specific activity of this waste is 10-3 — 10-4 Ci/litre.
A.J. BIRK: In the case of surface storage of bituminized waste, did you
take into account the fire risk involved? Were any precautions taken against
this hazard? What does you safety analysis say with regard to fire accidents?
E. MALÁáEK: The possibility of bitumen fire at the disposal site was
evaluated and as a result it was decided to pack the bituminous product into
metal barrels instead of cardboard drums, which had been originally proposed.
Furthermore, the volume of the bituminous product in the barrel was limited.
Tests have confirmed that fire in the vicinity of the barrels would not damage them.
For these reasons, fire protection is provided only in the bituminization plant.
K.G. ERIKSSON: I understand that your repository is designed also for
medium-level waste. As you have concrete structures and a drainage system,
surveillance and monitoring must be carried out over a period of several hundred
years. Are such long-term surveillance and monitoring requirements for a reposi
tory considered acceptable in Czechoslovakia?
J. KORTUS: Surveillance is proposed for a period of 240 years but experience
will show what frequency of sampling and period of monitoring is necessary.
R. KÔSTER: You describe the different barriers forming your containment.
For the leach rate of the product you give the value of 10"5 -10-6 g-cm-2 ' d-1.
In the case of PWR waste, especially ion-exchange resins, I would expect the leach
rates for your bitumen products to be higher by a factor of 10, owing to swelling
effects of the product in water.
IAEA-SM-243/156 219
J. KORTUS: Bituminization of spent higher-activity resins is only a temporary
solution. Laboratory-scale tests have been carried out on inorganic polymers
and an industrial process will soon be available. In this way the leach rate will
be reduced significantly. For safety reasons, however, we are still considering
an average leach rate of 10_s — 10-6 g'em '2 -d_1. Besides, the quantity of spent
resin is relatively small in comparison with other waste.
R. KOSTER: What is the difference between your storage concept and
that of “Eurostorage” at Mol in Belgium?
J. KORTUS: They differ in several respects. Our storage trenches are
located below ground level and therefore the drum transport system is arranged
in a different way. The specific activities of the waste are lower and there are
more concrete ponds. Our concept is much closer to the French or the Canadian
solution.
iSHALLOW LAND BURIAL: EXPERIENCE AND
DEVELOPMENTS AT LOS ALAMOS
J.L. WARREN
Los Alamos Scientific Laboratory*,
Los Alamos, New Mexico,
United States of America
] , IAEA-SM-243/150
Abstract
SHALLOW LAND BURIAL: EXPERIENCE AND DEVELOPMENTS AT LOS ALAMOS.
Since the mid-1940s, in excess of 170 000 m3 of low- and intermediate-level radioactive
solid waste, generated in operations at the Los Alamos Scientific Laboratory (LASL), has been
disposed of by on-site shallow land burial and retrievable storage in dry volcanic tuff. Guidelines
have been developed at LASL which regulate the construction of waste disposal facilities, burial
and storage operations, disposal site maintenance and restoration, and documentation of all
waste disposal activities. Monitoring programmes at the past and current solid waste disposal
sites have continued to show that, with the exception of low levels of tritium, no migration of
contaminants away from their disposal location has been detected.
INTRODUCTION
The Los A lamos S c i e n t i f i c L a b o ra to r y (LA SL ), o p e r a te d u n de r c o n t r a c t t o th e US D epa rtm en t o f E ne rg y , h as used o n - s i te s h a l lo w la n d b u r i a l as th e means to d is p o s e o f m ost low - and in t e r m e d ia t e - le v e l s o l i d r a d io a c t iv e w as te s g e n e r a te d a t LASL s in c e th e b e g in n in g o f o p e r a t io n s in th e m id - 1 9 4 0 's . The LASL, s i t u a t e d in th e s e m i- a r id s o u th w e s te rn r e g io n o f the U n ite d S t a t e s , h as v e ry f a v o r a b le c l im a t o lo g i c a l and g e o lo g ic a l c o n d i t io n s f o r d i s p o s a l o f s o l i d r a d io a c t i v e w as te s in t h i s m anne r . To d a t e , o v e r 170 ,000 m^ o f s o l i d r a d io a c t iv e w a s te has been b u r ie d m o n - s ite l o c a t i o n s . A d d i t i o n a l l y , s in c e 1971 , 1800 o ft r a n s u r a n ic (TRU) w as te has Seen r e t r i e v a b ly s to r e d .
T hrough December 31 , 1977 , th e r a d io a c t iv e c o n te n t o f b u r ie d w as te was an e s t im a te d 205 600 C i (d ecay c o r r e c te d ) , 94.6% o f w h ich was t r i t i u m . S to re d w astes c o n t a in 76 300 C i o f TRU a lp h a a c t i v i t y .
A l l i n v e s t i g a t i o n s o f th e s o l i d w as te d is p o s a l s i t e s have in d ic a t e d t h a t , w ith th e e x c e p t io n o f r e l a t i v e l y s m a ll q u a n t i t i e s o f t r i t i u m , no m ig r a t io n
* Work performed under the auspices of the US Department of Energy.
221
2 2 2 WARREN
LOS A LA M O S , N M
WESTMEAN ANNUAL PRECIP -15 inches.
______ 380 mm.
-MAIN AQUIFER
□ TUFF Ш a llu v iu m Ш BASALT Ш CONGLOMERATE ЕПШ SEDIMENTS Ш PERCHED WATER
EAST
PIEZOMETRIC SURFACE IN MAIN AQUIFER
APPROX. 3 MILES (5 km)
FIG.l. Geologic cross-section, Los Alamos, New Mexico.
o f r a d io n u c l id e s has been d e t e c t a b le . T h is p a p e r de s c r ib e s w as te d i s p o s a l f a c i l i t i e s , p r o c e d u r e s , exp e r ie n c e and d e v e lo p m e n ts a t LASL w h ich c o n t r ib u t e to t h i s s u c c e s s f u l p ro g ram . c
ENVIRONMENTAL SETTING
G e n e r a l S i t e E n v iro n m e n t
The LASL i s lo c a te d in n o r th c e n t r a l New M ex ico on th e P a j a r i t o P la t e a u , a t o p o g r a p h ic a l h i g h , ap p r o x im a te ly 600 m above th e R io G rande ( F ig . 1 ) . The
p l a t e a u i s c u t by num erous s te e p - s id e d canyons t h a t open i n t o th e R io G rande d r a in a g e a t i n t e r v a l s o f one to s e v e r a l k i lo m e t e r s . The L a b o r a t o r ie s ' p a s t andp r e s e n t d is p o s a l s i t e s a re lo c a te d on th e to p s o f th e
r e s u l t i n g f i n g e r - l i k e m esas . The p la t e a u i s composed o f m o d e ra te - to - d e n s e lv w e lded r h y o l v t i c t u f f s t h a t were d e p o s ite d a b o u t 1 .4 m i l l i o n y e a rs a g o . The t u f f i s b ro k e n by num erous n e a r - v e r t ic a l j o i n t s a t i n t e r v a l s o f 1 t o 2 m. The r e g io n a l w a te r t a b le i s l o c a t ed in s e d im e n ts o f th e R io G rande b a s in a t d e p th s o f 200 t o 300 m be low th e p la t e a u s u r f a c e . S o i l c o v e r
IAEA-SM-243/150 223
i s t y p i c a l l y <1 ш t h i c k , and w a s te - d is p o s a l f a c i l i t i e s a re dug th ro u g h th e s o i l i n t o th e u n d e r ly in g t u f f [1]. The Los A lam os a re a has an a v e rag e a n n u a l p r e c i p i t a t i o n w h ich v a r ie s w ith e le v a t io n from 300 t o 500 mm; th e p o t e n t i a l é v a p o t r a n s p i r a t io n i s ~ 600 mmin th e a re a [2]. The t u f f has a r e l a t i v e l y low hy d r a u l i c c o n d u c t iv i t y w h e n s a t u r a t e d , and t h i s de
c re a s e s r a p id ly w ith re d uce d m o is tu r e c o n t e n t . The com b ined e f f e c t o f th e p r e c i p i t a t i o n / e v a p o r a t i o n r a t i o and th e r e l a t i v e im p e r m e a b i l i t y o f th e t u f f r e s t r i c t s m o is tu r e p e n e t r a t i o n to th e u p p e r few m e te rs o f th e t u f f . D a ta o b ta in e d th ro u g h a n a ly s e s o f t u f f removed in c o r in g , and th ro u g h n e u t r o n m o is tu r e lo g g in g , show t h a t th e m o is tu r e c o n te n t o f th e t u f f i s g e n e r a l l y 5% o r l e s s , by v o lu m e , a t d e p th s >4 m [3-4] .
A c t iv e D is p o s a l S i t e
The m a jo r c u r r e n t LASL d is p o s a l s i t e , i d e n t i f i e d
as A rea G ( F ig . 2) has been in o p e r a t io n s in c e 1957. A rea G i s lo c a te d on an a p p ro x im a te ly - 3 .2 km lo n g mesa w h ich i s r e l a t i v e l y n a r r o w, r a n g in g in w id th "from a b o u t 90 to 400 m. The fe n ced p o r t io n o f th e s i t e c o v e rs 250 000 m^ o f th e e s t im a te d 360 000 m^a v a i l a b l e . To d a t e , a p p r o x im a te ly 135 000 m^ hasbeen u t i l i z e d . S o i l c o v e r on th e mesa i s 0 .3 to 0 .6 m t h i c k , and e r o s io n is s low b ecause o f th e s m a ll d r a in a g e a r e a . H e ig h t o f th e mesa above th e a d ja c e n t
canyon f l o o r s ra ng e s from 24 t o 30 m [5].
WASTE FORM AND PACKAGING
C h a r a c t e r i s t i c s o f B u r ie d W aste
Over th e p e r io d 1971-1978 , th e LASL has b u r ie d a n n u a l ly an a v e ra g e o f 6735 m3 o f r a d io a c t iv e s o l i d w a s te . A c tu a l a n n u a l vo lum es have ranged from 3610 m3 to 13075 m3 o v e r t h i s p e r io d . The ty p e s and r e l a t i v e q u a n t i t i e s o f th e LASL w as te s a re l i s t e d in T a b le I . The d i s t r i b u t i o n o f r a d io a c t i v e c o n ta m in a n ts fo und in th e s e w as te s is i d e n t i f i e d in T ab le I I . Over th e p e r io d 1975-1978 , 70.8% (231 13 .5 m3 ) o f th e w aste b u r ie d has o r ig in a t e d from f a c i l i t y decom
m is s io n in g and a re a d e c o n ta m in a t io n o p e r a t io n s , w h i le o n ly 29.2% (9 5 2 8 .5 m3 ) h as o r ig in a t e d from l a b o r a t o r y o p e r a t io n s . The b u r i a l vo lum e o f r o u t in e l a b o r a t o r y
гоN>4
legend
. PIT EXCAVATEO
- PIT UNEXCAVATED
PIT FILLED
HORIZONTAL
р^^г"'|Г"|Г*( MESA R 'M
FIG.2. Layout of Area G disposal/storage site.
WA
RR
EN
IAEA-SM-243/150
TABLE I. LASL BURIED WASTES (1971-1978)
225
Waste Cateqory Volume Volume %Laboratory Trash 13910 25.8Failed Equipment 2715 5.0Building Rubble 10625 19.7Sludge 1235 2.3Cement Paste 2755 5.1So il 19105 35.5Oil 120 0.2Uranium and Residues 260 0.5Filter Media 115 0.2Hot Cell Waste 40 0.1Graphite 535 1.0Animal Tissue 30 0.1Chemical Wastes 210 0.4Other 2225 4.1
Total 53880 100.0%
w aste has been re d uce d s i g n i f i c a n t l y as a r e s u l t o f em p loyee e d u c a t io n p ro g ra m s , im proved m a t e r ia ls h and l i n g , w r i t t e n p r o c e d u r e s , and w as te t r e a tm e n t p r i o r to b u r i a l .
W aste C o m pac tio n
S in c e A p r i l 1977 , a 45 500 kg p re s s com pac to r- b a le r has been used to re d uce th e vo lum e o f t r a s h - ty p e w as te s p r i o r to b u r i a l ( F ig . 3 ) . A p p ro x im a te ly 1000 m- o f w as te s a n n u a l ly a re re d uce d t o a b u r i a l vo lum e o f <200 m ^. B a le s g e n e ra te d a re 0 .4 m^ in vo lum e and w e igh 200 t o '¿Bu ксП A dv an ta g e s o f comp a c t io n in c lu d e a r e d u c t i o n i n f i r e h a za rd a s s o c ia te d w ith p l a c in g th e c o m b u s t ib le t r a s h w as te s d i r e c t l y i n t o a b u r i a l p i t , and a c o n s id e r a b le s a v in g o f p i t space due b o th t o th e d i r e c t r e d u c t io n in w as te vo lum e and to a r e s u l t a n t r e d u c t io n o f b a c k f i l l mat e r i a l r e q u ir e d to c o v e r t h i s s m a l le r w as te v o lu m e .
B u r ie d W aste P a c k a g in g
W aste p a c k a g in g in m ost in s t a n c e s se rv e s to m eet r e q u ir e m e n ts o f s a f e , o n - s i t e h a n d l in g and t r a n s p o r t . T y p ic a l p a c k a g in g s in c lu d e p l a s t i c w rap and b a g s , c a rd b o a rd b o x e s , m e ta l o r f i b e r d ru m s , and wooden c r a t e s . Large e q u ip m e n t ite m s and much o f th e
TABLE II. LASL WASTE RADIOACTIVfl'CONTAMINANTS (1971-1978)
226 WARREji j
Radionuclide Volume m-3 Volume %
Transuranics( 238P u , 2 39 pUf 2 41AItlf 23 3U}
35155 65.3
Uranium(depleted, normal, enriched)
14190 26.3
Fission Product/Induced Activity 3675 6.8Tr i t i urn 460 0.9Other 400 0.7
Total 53880 100%
decontamination/decommissioning wastes are not p a c k aged, but are delivered to the burial site in covered or enclosed transport vehicles. Following burial, the dry site environment contains the waste and its radioactive contamination.
T r i t iu m W aste
A n o ta b le e x c e p t io n to th e above p a c k a g in g and c o n ta in m e n t r e q u ir e m e n ts i s t r i t i u m c o n ta m in a te d w a s te . T r i t iu m i s h ig h ly m o b ile even u nde r d ry
b u r i a l s i t e c o n d i t i o n s . Found in w aste p r i n c i p a l l y as t r i t i a t e d w a te r , t r i t i u m i s as c a p a b le o f m ig r a t i o n as th e w a te r w h ich e x is t s in th e t u f f . LASL r e s e a rc h e f f o r t s in 1970 i d e n t i f i e d a s p h a l t as b e in g e f f e c t i v e in c o n t a in in g t r i t i u m c o n t a m in a t io n [6]. C o n s e q u e n t ly , a l l w as te s h a v in g >mCi l e v e l s o f t r i t ium r e q u ir e c o n ta in m e n t in m e ta l drums h a v in g an i n t e r n a l a s p h a l t c o a t in g . W aste c o n t a in in g t r i t i u m in excess o f a few 1 0 's o f c u r ie s r e q u ir e s c o n ta in m e n t in a m e ta l drum t h a t i s c o m p le te ly encased by s e v e r a l c e n t im e te r s o f a s p h a l t w i t h i n a second l a r g e r drum .
T r a n s u r a n ic (TRU) W aste
In March 1970 th e US F e d e r a l Governm ent is s u e d a r e q u ir e m e n t t h a t a l l w as te c o n ta m in a te d w ith lo ng -
l i v e d TRU a lp h a e m it t e r s a t an a c t i v i t y c o n c e n t r a t io n g r e a t e r th a n 10 n C i/ g m ust be s to r e d r e t r i e v a b ly in a c o n t a m in a t io n - f r e e m anner f o r a 20-year p e r io d .
IAEA-SM-243/150 227
FIG.3. Waste compactor-baler at Area G.
These TRU w as te s c o m p r is e 10 t o 15% o f LA SL 's a n n u a l w aste g e n e r a t io n . W astes in c lu d e e s s e n t i a l l y a l l mat e r i a l s l i s t e d in T ab le I ; TRU c o n ta m in a n ts a t LASL in c lu d e p r im a r i l y 238^239pu> .¿41дт у and ^ j U.
T r a n s u r a n ic w as te s r e c e iv e s p e c ia l h a n d l in g and a re packaged o n ly in c o n t a in e r s m e e tin g s t r i c t q u a l i t y c o n t r o l r e q u ir e m e n ts . These p a c k a g in g s in c lu d e 210-£ drums and f i b e r g la s s - r e in f o r c e d p o ly e s t e r c o a te d wooden c r a t e s o f v a r y in g s i z e . A l l p a ck ag e s in c lu d e in n e r p l a s t i c l i n e r s , r a n g in g in t h ic k n e s s from 0 .1 3 mm f o r d ry w as te s to a 2 .3 mm h ig h d e n s i t y c r o s s l in k e d p o ly e th y le n e l i n e r f o r p o t e n t i a l l y c o r r o s iv e w a s te s .
DISPOSAL FACILITY CONSTRUCTION AND USE
C r i t e r i a
In 1974 , LASL W aste Management and E n v ir o n m e n ta l s c i e n t i s t s s i g n i f i c a n t l y expanded c r i t e r i a , d e v e lo p e d
228 WARREN
FIG.4. Waste burial pit measuring 180 m long, 15 m wide, 8 m deep.
i n 1965 by th e US G e o lo g ic a l S u rv e y , t o e n su re ade q u a te c o n ta in m e n t o f b u r ie d r a d io a c t iv e w as te a t LASL s i t e s . The c r i t e r i a a p p ly c o n t r o ls to c o n d i t io n s r e l a t i n g t o p o s s ib le pa thw ays o f r e le a s e o f c o n ta m in a t i o n from b u r ie d w a s te s . M a jo r r e le a s e pa thw ays a re th ro u g h i n f i l t r a t i o n o f w a te r i n t o w a s te s , and e r o s io n , b o th from th e s u r fa c e and from th e mesa s id e s . The c r i t e r i a d e s c r ib e d is p o s a l f a c i l i t y s i t e s , o r ie n t a t i o n , c o n s t r u c t io n , a p p r o v a l , u s e , d o c u m e n ta t io n , m a in te n a n c e , and s i t e - c o n d i t i o n i n g r e q u ir e m e n ts f o l lo w in g u s e .
B u r i a l P i t C o n s t r u c t io n
The b u lk o f LASL w as te is b u r ie d in la r g e p i t s ( F ig . 4 ) . E xcep t f o r p i t d e p th , p i t d im e n s io n s gen e r a l ly - a re n o t r e s t r i c t e d , and have rang e d in s iz e f rom 120 t o 180 m Io n a bv 8 to 30 m w id e . To p re v e n t p o s s ib le a s s o c ia t io n betw een b u r ie d w as te and pe rched w a te r w h ich e x is t s in a l l u v i a l m a t e r i a l in th e f lo o r s o f a d ja c e n t c a n y o n s , no b u r i a l f a c i l i t i e s may be d e e p e r th a n th e a d ja c e n t canyon f l o o r . To d a te th e
IAEA-SM-243/1 SO 229
d e e p e s t p i t e x c a v a t io n has been - 1 4 m. t h i s b e in g
c o n t r o l l e d p r im a r i l y By t h e s t a b i l i t y o f th e p i t w a l l s . P i t s a re o r ie n te d w ith th e lo n g d im e n s io n as p a r a l l e l as p o s s ib le to th e a re a s u r f a c e c o n to u r s . P i t s a r e , dug no c lo s e r th a n 15 m to a canyon w a l l , and p i t s id e w a l ls m ust be a t l e a s t 4 .5 m a p a r t a t th e s u r f a c e . A l l t o p s o i l i s removed and s t o c k p i l e d f o r f u t u r e s i t e r e v e g e t a t io n .
For s t a b i l i t y , p i t w a l ls a re e x c a v a te d w ith ap p r o x im a te ly a 1 :4 s lo p e . The ends a re dug w ith s lo p e s r a n g in g from 2 :1 t o 4 :1 to a l lo w a c c e s s by ve
h i c l e s and e q u ip m e n t . As a f i n a l s te p in th e excav a t i o n , t u f f i s g round and com pacted in th e p i t b o t tom t o a d e p th o f 0 .1 5 t o 0..10~пй T h is p r o v id e s a s e a l f o r f r a c t u r e s in th e p i t b o tto m , and an a b s o rp t i o n medium f o r p r e c i p i t a t i o n t h a t e n te r s th e p i t p r i o r to w as te b u r i a l . A l l b u r i a l f a c i l i t i e s a re su rvey e d and re co rd e d on pe rm anen t LASL d r a w in g s . LASL e n v ir o n m e n ta l s c i e n t i s t s map a l l f r a c t u r e s in p i t w a l l s , and a l l open f r a c t u r e s in ex ce ss o f 5 cm w ide a re f i l l e d w itn m a t e r ia l s such as c em e n t, ben t o n i t e c la y , o r c ru she d t u f f . F i n a l a p p r o v a l o f a p i t t o r use by a LASL g e o lo g is t is r e q u i r e d .
P i t B u r i a l O p e r a t io n s
W aste i s b u r ie d in p i t s in l a y e r s , w ith a m in i mum o f 0 .1 5 m o f b a c k f i l l com pacted b e tw een la y e r s . E x cava ted t u f f i s used f o r a l l b a c k f i l l o p e r a t io n s . W aste c o n s id e re d to be c o m b u s t ib le , w ind d i s p e r s a b le , o r w h ich has a p o t e n t i a l f o r c o n t a m in a t io n r e le a s e m us t be cove red on th e day o f d e l i v e r y . R o u t in e ly , w as te i s cove red tw ic e w e e k ly . P i t s a re f i l l e d to a l e v e l 1 m be low th e " s p i l l p o i n t , " d e f in e d a s th e lo w e s t p o in t on th e p i t r im , th u s e n s u r in g co m p le te c o n ta in m e n t o f w as te by u n d is tu r b e d t u f f . Based on c o n s e r v a t iv e e s t im a te s o f r a te s o f s u r fa c e and mesa w a l l e r o s io n , t im e s to expose b u r ie d w as te s from the to p and s id e s a re e s t im a te d to be 50 000 and 100 000 y e a r s , r e s p e c t i v e l y . f i ] In r e c e n t y e a rs th e w aste vo lum e b u r ie d in a p i t has ranged from 18 to 42%. w ith th e av e rag e b e in g 31% o t th e t o t a l p i t v o lum e .
F in a l Cover and S i t e R e h a b i l i t a t i o n
E xcava ted t u f f i s used to r e f i l l th e p i t to th e o r i g i n a l la n d c o n to u r . A d d i t i o n a l m ound ing w ith t u f f
' З » / 6t z f f e X -
230 WARREN
TABLE III. RE VEGETATION SEED MIXTURE FOR SEMI-ARID NEW MEXICO
Indian Rice Little Blue-stem Western Wheatgrass Side Oats Grama Yellow Blossom Clover Buffalo Sharps Perennial Rye
(Oryzopeis hymenoides) (Andropagan scoparius) '(Ag ropy ron Smithii) (Boutelana curtipendula) (Melilotus oficinalis) (Buchloe dactyliodes) (Lelium perenne)
UNLINED LINED STORAGE
TYPICAL SHAFT DIAGRAM
TYPICAL ORIENTATION FOR О.З-О.Э-m-DIAMETER SHAFTS
FIG.5. LASL burial/storage shafts.
IAEA-SM-243/150 231
up t o 1 m o r more a t th e c e n te r o f th e p i t , and e x te n d in g to 1 m beyond th e edges o f th e p i t i s a c c o m p lis h e d i f r e q u ir e d f o r s u r fa c e d r a in a g e mod i f i c a t i ô n .
F o l lo w in g th e f i n a l use o f an a r e a , r e v e g e t a t io n i s r e q u ir e d . T o p s o il i s sp re ad to a d e p th o f 10 t o 15 cm, fo l lo w e d by a p p l i c a t i o n o f s e e d , f e r t i l i z e r ,
and s tra w ( t o m a in t a in s o i l m o is t u r e ) . To m eet th e s e m i- a r id g ro w ing c o n d i t io n s a t Los A lam os a seed m ix tu r e c o m p r is ed o f a p p r o x im a te ly e q u a l p r o p o r t io n s o f th e seven v a r i e t i e s l i s t e d in T ab le I I I a re u se d . S p e c ia l w a te r in g i s n o t r e q u ir e d .
B u r i a l S h a f t C o n s t r u c t io n and Use
To p r o v id e f o r b e t t e r i s o l a t i o n f o l lo w in g b u r i a l a n d /o r to in c r e a s e w orker s a f e t y , c e r t a i n LASL w as te s a re b u r ie d in deep s h a f t s ( F ig . 5 ) . These w as te s and s p e c i f i c re a s o n s f o r s h a f t b u r i a l a re l i s t e d in T ab le IV . B u r i a l s h a f t s a re auge red v e r t i c a l l y i n t o t h e t u f f , m ost a re u n i in e d Га few w e r e c o n s t r u c t e d w ith a0 .3 m c o n c re te l i n e r ) , and m easure 0 .3 to 1 .8 m in d ia m e te r by up to 20 m d e e p . A l l a p p l i c a b le d is p o s a l f a c i l i t y s i t i n g and c o n s t r u c t io n c r i t e r i a d e s c r ib e d f o r b u r i a l p i t s a p p ly to s h a f t s .
To a s s u re c r i t i c a l i t y s a f e t y , s h a f t s a re l im i t e d t o 500 g t o t a l f i s s i l e m a t e r i a l . S h a f t s may be f i l l e d to w i t h in ~ 1 .5 m o f th e s u r f a c e . S u f f i c i e n t e x c a v a te d t u f f th e n i s added to c o v e r a l l w as te in th e s h a f t , and a f i n a l 1 m c o n c re te p lu g i s p o u re d .
TRANSURANIC WASTE STORAGE
T ra n s u ra n ic w as te s a re s to r e d in f a c i l i t i e s d e s ig n e d from m o d i f i c a t i o n o f b u r i a l p i t s and s h a f t s . The b u lk o f th e LASL t r a n s u r a n ic w astec u r r e n t ly i s s to r e d in a m o d i f ie d p i t ( F ig . 6)m e a s u r in g 122 m lo n g , 9 m w ide and 8 m d e e p . p i t
m o d i f i c a t i o n s in c lu d e th e n e a r- v e r t i c a l e x c a v a t io n o f one e n d , pavem ent o f th e p i t f l o o r w ith a s p h a l t ro ad m a t e r i a l , ' and c o n s t r u c t io n o f two 1 . 2 m d ia m e te r , 3 m deep a s p h a l t - 1 in e d sumps f o r c o l l e c t i o n and m o n ito r in g o f p r e c i p i t a t i o n .
W aste p ackages (drum s and c r a t e s ) a re s ta c k e d on th e p ave m e n t, b e g in n in g a t th e v e r t i c a l e n d , to a
232 WARREN
TABLE IV. LASL BURIAL SHAFT USAGE
Wastes ____________________ Reason for shaft use1. High Activity ,
(>mCi quantities)
2. Beta-gamma hot cell waste (intermediate level)
3. High-activity accelerator waste
4. Contaminated chemical wastes
5. Animal tissue
6. Cement paste
Greater protection of packaging; ease of monitoring.Direct disposal from bottom loading, truck-mounted cask; personnel protection.Personnel protection.
Greater protection of packaging; improved isolation from other reactive wastes.Isolation from meat-eating animals.Ease of operation; paste is pumped directly into shaft from adjacent liquid-waste treatment facility (Note:This operaton is not at Area G) .
h e ig h t o f 6 to 7 m. As th e s ta c k p ro g re s s e s down th e p i t , th e to p is cove red w ith 19 mm t h i c k p ly w o o d , and
th e e n t i r e s ta c k is e ncased w ith 0 .5 nun n y lo n r e in f o r c e d v in y l s h e e t in g . The s ta c k is cove red and mounded w ith t u f f in th e same m anner as a b u r i a l p i t .
Three d i f f e r e n t s to r a g e modes a re u t i l i z e d f o r o t h e r s p e c ia l t r a n s u r a n ic w a s te s ; th e se w as te s and s to r a g e modes a re i d e n t i f i e d in T ab le V (see also Figs 7,8).
WASTE RECORDS AND FACILITY IDENTIFICATION
Perm anent re c o rd s f o r a l l w as te s b u r ie d and s to r e d s in c e 1971 a re c o n ta in e d in a com pu te r re c o rd s s y s te m . F ig . 9 is th e d a t a i n p u t fo rm u se d . Ite m 5- WASTE CODE - i s a code used to i d e n t i f y th e w aste m a tr ix ( e .g . c o m b u s t ib le t r a s h , s lu d g e , b u i l d i n g r u b b l e , e t c . ) . DISPOSAL/STORAGE LOCATION - ite m 14 - in c lu d e s e n t r ie s f o r th e l o c a t i o n o f w as te w i t h i n a p i t . POST(S) r e f e r s to s e q u e n t i a l l y num bered m arke rs
IAE A-SM-243/150 233
FIG.6. TRU-waste storage pit. i Z Z J i ^
a t 3 m i n t e r v a l s a lo n g th e p i t s id e , P i t LAYER is num bered s e q u e n t i a l l y b o tto m t o t o p , and POS. ( P o s i t i o n ) d e n o te s th e s id e o r c e n te r p o r t io n o f th e p i t .
When f i l l e d , and f o l l o w in g r e v e g e t a t io n o f th e a r e a , p e rm anen t b ra s s cap m a rke rs a re s e t in c o n c re te m onuments a t two c o rn e r s o f e ach p i t . M arkers id e n t i f y th e p i t a n d 'p i t c o r n e r s , d a te s o f u s e , and m a jo r c o n t a m in a n t s . S im i l a r b r a s s m a rke rs a re s e t in th e c o n c r e te cap s e a l in g each b u r i a l s h a f t .
SITE MONITORING
R o u t in e M o n ito r in g
The o v e r a l l L a b o ra to r y s i t e e n v ir o n m e n ta l s u r v e i l l a n c e prog ram a t LASL i s n o t s u f f i c i e n t l y s e n s i t i v e , w ith th e p o s s ib le e x c e p t io n o f a tm o s p h e r ic t r i t iu m , to d i s t i n g u i s h a b u r i a l g ro u n d r e le a s e from n o rm a l s i t e r e le a s e s . C o l l e c t i o n o f w a te r sam p le s from u n s a tu r a te d t u f f i s d i f f i c u l t to im p o s s ib le , and
234 WARREN
TABLE V. TRANSURANIC WASTE STORAGE MODES
Waste Storage Mode238 Pu-contaminated trash, residues, small equipment; >1 g 238Pu per drum.Highly beta-gamma active hot-cell wastes; up to several 1000's R/h per 4- package, unshielded.
239 Pu _ 241Am cementpaste.
1 1 0 - I steel drums, 2 each sealed in buried concrete cask (Fig. 7). '
1 2 packages sealed in 0.6 m diameter, 4.5 m long corrugated metal pipe/concrete cask set in 0.9 m diameter, 4.5 m deep shaft (Fig. 5).Paste solidified in 0.75 m- diameter, 20 m-long corrugated metal pipe having 0.3 m-thick uncontaminated concrete plug at the bottom. Pipe filled to within 0.3 m of top, then sealed with uncontaminated concrete. Pipes stand vertically in 6 m-deep pit; backfilled with 1 m tuff (Fig. 8 ) .
c o n s e q u e n t ly sam p le s o f t u f f w ith t h e i r c o n ta in e d wat e r a re a n a ly z e d . W ate r s am p le s o b ta in e d from n ea rb y s u p p ly w e l ls a re a n a ly z e d f o r r a d io n u c l i d e s .
S o i l m o is tu re c o n te n t and movement is d e te rm in e d to d e p th s o f 40 m in u n d is t u r b e d t u f f and in and be low f i l l e d b u r i a l / s t o r a g e p i t s . 'M é té o r o lo g ie d a ta c o l l e c t e d a t a b u r i a l s i t e f a c i l i t y p r o v id e s a d d i t i o n a l in fo r m a t io n on th e r a t e s o f w a te r movement i n to and o u t o f th e t u f f . From th e s e d a ta an u ppe r l i m i t f o r th e r a te o f p o s s ib le r a d io n u c l id e m ig r a t io n from w a te r movement is b e in g d e te rm in e d .
Low le v e ls o f TRU, f i s s i o n p r o d u c t , and t r i t i u m c o n ta m in a t io n have been d e te c te d in a n a ly s e s o f s u r fa c e s o i l sam p les c o l l e c t e d a t two o ld LASL b u r i a l
s i t e s . W h ile t h i s c o n t a m in a t io n , w ith th e p o s s ib le e x c e p t io n o f t r i t i u m , i s b e l ie v e d to be th e r e s u l t o f o p e r a t io n a l " s p i l l s " w h i le th e s i t e s were a c t i v e , a d d i t i o n a l s a m p lin g and i n v e s t i g a t i o n is p la n n e d .
IAEA-SM-243/150 235
T YP IC A L CASK BU R IAL
-0.9 ■ ,
H! •- —1.5 «-|-1.г «Н o.s --4-i.s»—H.!«H ^
Ц «4-1.5 « —j
У---09 ■
C ASK LAYOUT DETAIL
FIG. 7. Concrete storage cask usage.
C o r in g
In 1976 , f i v e h o r i z o n t a l c o re h o le s were d r i l l e d b e n e a th P i t 3 in a fa n - s h a p e d a r r a y . (See F ig . 2; P i t 3 r e c e iv e d w aste 1 9 6 3 - 1 9 6 6 .) [7]. The h o le s a n g le d s l i g h t l y dow nw ard , and d e p th s ranged from ze ro to seven m e te rs b e n e a th th e p i t . C ores were a n a ly z e d f o r q ro s s - a , g r o s s - 6 , ^ S r , l 3 7 ^ Sf t o t a l u r a n iu m , 2 38 f Z39pUf and ^ 4 l ^ # jj0 s t a t i s t i c a l l y s i g n i f i c a n t
v a r i a t i o n s o f a c t i v i t i e s were o b se rv e d betw een th e se
c o re s and sam p le s c o l l e c t e d in th e same g e o lo g ic s t r a t a a t o th e r l o c a t i o n s [8]. T r i t iu m was n o t
236 WARREN
FIG. 8. Corrugated metal pipe for storage of cement paste.
a n a ly z e d f o r because a i r had been used as a c u t t i n g s c a r r i e r d u r in g c o r in g , and th u s w ou ld have removed an unknown am ount o f w a te r v a p o r c o n t a in in g t r i t i u m . No r a d io n u c l id e s were d e te c te d in th e c o re s t h a t c o u ld be a t t r i b u t e d to m ig r a t io n from th e o v e r ly in g w a s te .
T r i t i u m M i g r a t i o n
T r it iu m m ig r a t io n was d e te c te d in 1970 , and th e use o f a s p h a l t was i n i t i a t e d th e n to a c h ie v e b e t t e r c o n ta in m e n t . M o n ito r in g o f t r i t i u m d is p o s a ls th ro u g h 1975 , how ever, showed t h a t a s p h a l t , as i t was b e in g a p p l ie d to s h a f t w a l ls and as an in n e r drum c o a t in g , was i n e f f e c t i v e in e n h a n c in g c o n ta in m e n t [9]. R e s u l t s from two b u r i a l l o c a t io n s showed t h a t <1% o f th e b u r ie d t r i t i u m was m ig r a t i n g , and t h a t th e m ig r a t io n was r e s t r i c t e d to r e l a t i v e l y s m a l l a re a s im m e d ia te ly a ro und th e d is p o s a l l o c a t i o n s . F u r t h e r , i t was shown t h a t th e t r i t i u m , as t r i t i a t e d w a te r , was m ov ing w ith
IAEA-SM-243/150 237
PLEASE READ INSTRUCTIONS ON BACK CAREFULLY H' 7 Waste Management
LASL RADIO ACTIVE SOLID WASTE EXt 6095 MS' 592DISPOSAL RECORD FORM
4. O R IG IN OF WASTE ОzSGROUP TA BLOG. ROOM
i 1 1 1 i
5. WASTE COOE
i i1 2 - 8 9 14 16 20 21 25 26 27 28 32 33 34 37 38 40
6. WASTE DESCRIPTION
41M - MITIN F - FKKT>
> 80
<1 Al 1 О
PLASTICBAGS
CARDBOARDBOXES
DRUMS WOODEN CRATES 8. GROSS eSURFACE
MR/HR1 METER MR/HRNO. GAL. NO. VOLUME - ft3 VO LUM E z
Э
2| i i ' 1 i 1 ' 1- 1 ' ' I ' , , 1 . . . ? . , » . ... i1 i i i i 111 12 14 15 17 18 19 20 21 22
■ KILOG RAM26 27 30 31 32
10. GROSS T - TON
W EIGHT zЭ 11. A D D IT IO N A L DESCRIPTION OF PACKAG IN G A N D PAC KAG IN G M A T E R IA LS
. . T 1 1 1 1 1 1 1 1 1 J i 1 1 1 1 - 1 1 1 i 1 1 i i 1 1 1 1 1 1 1 ! 1 142 45 46 47
i1 2 - 8 9
12. R A D IO N U C LID E CONTENT С - CURIE M- QRAM AMOUNT DETERMINED BY:
SS MATERIALS»f Л1 1
NUCLIDE ■ AMOUNT ± zЭERROR ON AMOUNT *1
M - M IA IU N IM IN T К - ЖЖТ1МАТС ACCOUNTPROJECT
CODE
f ,E
T .E
i
t !E t I E
t ,E
? .E
t , ,E , T . 1
Ei
WASTE GENERATOR H-1 A R EA REPRESENTATIVE GROUP LEAD ER (AS NECESSARY)S igna ture c e r t i f ie s th a t waUe is in accordance S ignature c e r t i f ie s th a t waste package orwith »11_agpHc<ble disposal r«quireoents. sMpaent I t *af« to handle >nd tr inspo rt.
013. D ATE DISPOSED* m о о л
I I I 1-
14. D ISPO SAL/STO RAG E LO CATIO N
AREA SHAFT PIT POST(S) LAYER i
14 15 16 17 19 20 21 22 26 26 27 26 29
15. SHAFT SURFACE DOSE
MR/HR
H-7 WASTE MANAGEMENT REPRESENTATIF
FIG.9. LASL radioactive solid waste disposal form.
238 WARREN
w a te r v a p o r th ro u g h j o i n t s and p o ro u s z o n e s , r a t h e r th a n w ith l i q u i d m o is tu r e f lo w . No o th e r r a d io n u c l i d e s p r e s e n t in LASL w as te can m ig r a te in t h i s
f a s h io n . S in c e 197 5 , p ro c e d u re s have been e s t a b l i s h ed t o a s s u re c o n ta in m e n t o f t r i t i u m u s in g a s p h a l t . B u r i a l o f t r i t i u m w as te has been a c c o m p lis h e d in a new p o r t io n o f th e b u r i a l s i t e in a s h a f t s u r ro u n d e d by n in e s a m p lin g h o le s . T h is s h a f t w i l l be s e a le d in m id- 1 979 , and s a m p lin g f o r t r i t i u m and s o i l m o is tu r e w i l l be i n i t i a t e d s h o r t ly t h e r e a f t e r .
SUMMARY
O ver 30 y e a rs o f e x p e r ie n c e show t h a t s h a l lo w la n d b u r i a l o f low- and in t e r m e d ia t e - le v e l s o l i d r a d io a c t i v e w aste can be s a f e ly a c c o m p lis h e d in an en v i r o n m e n t a l ly a c c e p ta b le m an n e r . F u tu re LASL W aste M anagement o p e r a t io n s w i l l s t r o n g ly em ph as ize d e v e l opm ent and im p le m e n ta t io n o f im proved t e c h n o lo g ie s in
th e f i e l d s o f w as te vo lum e r e d u c t io n , a r i d - s i t e m on it o r i n g , and o v e r a l l s h a l lo w - la n d b u r i a l p r o c e d u re s .
REFERENCES
[1] WHEELER, M. L . , SMITH, W. J . , GALLEGOS, A . F . , A P r e l i m i n a r y E v a l u a t i o n o f t h e P o t e n t i a l f o r P l u to n iu m R e le a s e from B u r i a l G rounds a t Los A lamos S c i e n t i f i c L a b o r a to r y , Los A l a m o s S c i e n t i f i c L a b o ra to r y r e p o r t LA-6694-MS, F e b ru a ry 1977.
[2] TUAN, Y . F . . , EVERARD, C . E . , W IDDISON, J . G . , BENNETT, I . , The C l im a te o f New M e x ico , New Mexic o P la n n in g O f f i c e (1 9 7 3 ) .
[3] T ra n s u r a n ic S o l id W aste M anagement P rog ram s : J u ly - December 1974 , Los A lam os S c i e n t i f i c Labo r a t o r y r e p o r t LA-6100-PR, O c to b e r 1975 .
[4] PURTYMUN, W. D . , KENNEDY, W. R . , G eo logy and Hyd r o lo g y o f M e s ita d e l B uey , Los A lam os S c i e n t i f i c L a b o ra to r y r e p o r t LA-4660 (1 9 7 1 ) .
[5] PURTYMUN, W. D . , G e o lo g y and H y d ro lo g y o f A reaG, M e s ita d e l B uey , Los A lam os C o u n ty , New Mexi c o , US G e o l. S u rvey a d m in i s t r a t i v e r e le a s e
(1 9 6 6 ) .[6] EMELITY, L . A . , CHRISTENSON, C . W. , WANNER, J .
J . , T r i t iu m Loss from C oa ted Cement P a s te B lo c k s , Los A lam os S c i e n t i f i c L a b o ra to r y r e p o r t LA-DC-12740 (1 9 7 2 ) .
IAEA-SM-243/150 239
[7] PURTYMUN, W. D . , WHEELER, M. L . , ROGERS, М. А . , G e o lo g ic D e s c r ip t io n o f C o res from H o le s P-3 MH-1 th ro u g h P-3 MH-5 A rea G , T e c h n ic a l A rea 54 , Los A lamos S c i e n t i f i c L a b o ra to r y r e p o r t LA-7308- MS (1 9 7 8 ) .
[8] N u c le a r W aste M anagem ent T e c h n ic a l D eve lopm en ts P ro g re s s R e p o r t , Jan uary - D ecem be r 1978 , Los A la mos L a b o ra to r y r e p o r t , in p r e s s .
[9] WHEELER, M. L . , WARREN, J . L . , " T r i t iu m C o n ta in m ent A f t e r B u r i a l o f C o n ta m in a te d S o l id W a s te ," P ro c e e d in g s o f 23rd C o n fe re n c e on Remote System s T echno logy (1 9 7 5 )1 0 0 .
DISCUSSION
E.G. RAMO: What is the reason for selecting a storage time of 20 years for
transuranic waste?
J.L. WARREN: The United States Government in 1970 established the
requirement for storage of transuranic solid wastes for a minimum period of
20 years. It is considered that during this time a facility for geologic disposal will
be developed for this waste.
H. KRAUSE: To what kind of pretreatment is the tritium waste subjected
prior to disposal? Is the tritiated water adsorbed or fixed in some material?
J.L. WARREN: Generally there has been no pretreatment of tritium waste.
Typical high-activity wastes include absorbed oils and tritiated water in molecular
sieves. Because of the high activities, which can range up to a few 1000 Ci/ltr,
no treatment is attempted. Instead, packaging is provided to contain the
contamination.
F.A. VAN КОТЕ: What rate of tritium migration was observed when tritium-
containing waste was contained by a layer of bitumen between two steel drums?
Does the migration rate increase with time?
J.L. WARREN : No migration of tritium has been observed at Los Alamos
from waste in this type of package. Specific leaching studies have shown little or
no leaching over a period of a few years.
J.B. ROBERTSON: Have you observed slumping and cracking of trench
caps which would allow rainwater to infiltrate into the buried wastes?
J.L. WARREN: Slumping and/or cracking of trench caps has been seen in
only a few instances at the currently active site. Any such occurrences are quickly
detected by routine monitoring of site conditions and corrective action eliminates
any possibility of water infiltrating in buried wastes. Present waste burial practices
of layering waste and compacted backfill in the pit generally preclude any
significant problems of this sort later on.
240 WARREN
Valentina BALUKOVA: Could you give more details about the packaging
of transuranic wastes?
J.L. WARREN: Transuranic solid wastes at Los Alamos generally contaii]
no significant fission product activity and thus packaging requires no shielding.
Metal drums and wooden crates having an outer coating of fibreglass are used
in most cases.
IAEA-SM-243/71
APPLICATION DE LA TECHNIQUE
DES BARRIERES CAPILLAIRES
AUX STOCKAGES EN TRANCHEES
D. RANÇON
CEA, Institut de protection et de sûreté nucléaire,
Centre d’études nucléaires de Cadarache,
Saint-Paul-lez-Durance, France
Abstract-Résu mé
APPLICATION OF THE CAPILLARY BARRIER TECHNIQUE TO STORAGE
IN TRENCHES.
The capillary barrier technique, by which one can create dry structures in porous media
relying simply on the differences of particle size distribution between two media, has been
demonstrated by laboratory experiments on reduced-scale models. In order to test the
applicability of this phenomenon to practical waste storage, an experimental trench was
excavated in fine soil, filled two-thirds with gravel and then covered with argillaceous sand
of fine texture which formed a dome above the soil surface. Moisture transport was observed
by measuring the evolution of water profiles with a neutron moisture gauge and by noting
the movement of radioactive tracers. Static measurements were also performed with a strainer
pipe going down to the bottom of the trench. The influence of atmospheric precipitation
was recorded during several seasonal cycles; experiments were also carried out with intensive
water spraying on the trench surface. After five years of regular observation, in spite of
the intensive irrigation conditions no appreciable moisture transport was observed across
the interface of the fine medium towards the coarse medium, which indeed remained a dry
structure. Applied with certain precautions, this simple, sturdy and inexpensive method can
considerably improve the safety of underground repositories for low- or medium-level wastes.
APPLICATION DE LA TECHNIQUE DES BARRIERES CAPILLAIRES AUX STOCKAGES
EN TRANCHEES.
La validité de la technique des barrières capillaires, qui permet de créer des structures
sèches en milieux poreux en opérant seulement sur les différences de granulométrie de deux
milieux, a été démontrée par des essais en laboratoire sur modèles réduits. Pour appliquer
ce phénomène au stockage des déchets, on a réalisé une tranchée expérimentale. Creusée
dans un sol fin, elle est remplie aux deux tiers par du gravier, puis comblée par un sable
argileux à texture fine qui constitue un dôme au-dessus de la surface du sol. Les transferts
d’humidité ont été observés par mesure de l’évolution des profils hydriques avec un humidi-
mètre à neutrons et par observation du transfert de traceurs radioactifs; on a aussi effectué
des mesures statiques au moyen d’un tube crépiné reposant sur le fond de la tranchée. On a
observé l’influence des précipitations atmosphériques pendant plusieurs cycles de saisons,
mais on a aussi effectué des expériences sous arrosage intensif de la surface de la tranchée.
241
Après 5 ans de contrôles réguliers et malgré des conditions rigoureuses d’irrigation, on n’observe
pas de transferts d’humidité sensibles à travers les interfaces du milieu fin vers le milieu grossier
qui constitue bien une structure sèche. En prenant certaines précautions lors de la mise en
oeuvre, ce procédé simple, robuste et peu coûteux doit permettre d’améliorer de façon notable
la sûreté des stockages par enfouissement des déchets de faible ou moyenne radioactivité.
242 RANÇÔJ4
1. INTRODUCTION: RAPPEL DU PRINCIPE DES
BARRIERES CAPILLAIRES
Dans le cadre d’études sur l’amélioration de la sûreté des stockages de déchets
radioactifs, on a montré, par des essais sur modèles réduits, qu’il était possible de
réaliser des structures sèches en milieux poreux, en agissant sur des barrières capil
laires créées par des différences de granulométrie, sans utiliser de barrières
imperméables.
Dans un modèle réduit constitué par une couche de sable fin surmontant
une couche de sable grossier, l’eau amenée sur la surface s’infiltre dans le sable
fin mais elle ne pénètre pas dans le sable grossier, quelle que soit la forme de
l’interface de séparation. Ce phénomène est dû à la prédominance des forces de
succion sur les forces de gravité. L’eau pénètre dans le sable grossier quand la
totalité du sable fin est humidifiée. On crée une structure sèche, même sous un
régime permanent d’infiltration non saturante, si le sable grossier est entouré de
tous les côtés par du sable fin. On a déterminé les granulométries limites au-dessus
desquelles il ne se produit plus.
Ce phénomène est décrit en détail dans le rapport [1] illustré par un film [2].
Le principe des barrières capillaires est schématisé sur la figure 1.
Ces expériences peuvent déboucher sur plusieurs applications dans le domaine
des circulations d’eau souterraines. On a considéré qu’il serait possible d’appliquer
cette technique à l’amélioration de la sûreté des stockages de déchets radioactifs.
D est usuel, quand les conditions hydrologiques et climatiques sont favorables,
d’enfouir les déchets dans des tranchées; la terre remblayée et tassée doit limiter
sans les empêcher les transferts d’eau aux alentours des produits stockés [3]. On
a donc étudié un procédé simple et peu coûteux destiné à améliorer cette technique
de stockage: dans la tranchée, les déchets seraient entourés d’un matériau grossier
dans lequel ils seraient maintenus à l’abri des eaux d’infiltration par l’intermédiaire
des barrières capillaires.
Pour tester ce procédé de stockage une tranchée expérimentale a été réalisée
sur le site du Centre d’études nucléaires de Cadarache en collaboration avec le
Service de contrôle des radiations du Centre.
IAEA-SM-243/71 243
FIG .l. Principe des barrières capillaires.
2. DESCRIPTION DE LA TRANCHEE EXPERIMENTALE
Cette tranchée a été implantée sur le site du CEN de Cadarache dans la zone
réservée à l’entreposage des déchets [3]. Ses dimensions sont montrées sur la
figure 2, c’est un modèle réduit à une échelle environ 1/2 des tranchées usuellement
utilisées.
La structure grossière est constituée par un gravier roulé de la Durance dont
les grains sont compris entre 10 et 15 mm.
La structure fine est constituée par le sol environnant (sol sablo-argileux,
fortement calcaire). Pour constituer la structure fine supérieure on n’a pas directe
ment utilisé la terre de déblai, agglomérée en motte, qui risquait de donner un
remplissage hétérogène avec chemins préférentiels pour les eaux d’infiltration;
on a utilisé une terre sablo-argileuse de même nature, préalablement écrasée, avec
des grains inférieurs à 0,2 mm.
Afin d’empêcher l’interpénétration de la terre fine dans les interstices du
gravier, une couche de laine de verre a été déposée à l’interface.
On a implanté dans cette tranchée deux tubes en alliage d’aluminium fermés
à leur base; ces tubes, destinés à recevoir la sonde d’un humidimètre à neutrons,
pénètrent dans le sol jusqu’à 1 mètre sous le fond de la tranchée (fig.3 et 4). Au
niveau des tubes les interfaces se trouvent respectivement à 75 et 250 cm sous la
surface du dôme de terre; l’interface supérieure a la forme d’un toit à versants
symétriques, ces tubes (Al et A2) étant situés au faîte.
COUPE A
COUPE В
T e rre Ппс ra p p o rté e de m ê m e n a tu re que la t e r re environnante, g ra in s < 0,2 m m
e° °0 c G ra v ie r I g ra in s e n tre 10 « t 15 m m )
. i - ^ E m p la c e m e n t des th e rm o ré sis ta n ce s
FIG.2. Plan et coupes de la tranchée expérimentale.
244 R
AN
ÇO
N
IAE A-SM-24 3/71 245
FIG.3. Vue de la tranchée avec la structure grossière.
Un troisième tube A3 a été implanté jusqu’à la même profondeur à l’extérieur
de la tranchée pour mesurer l’évolution de l’humidité dans le sol environnant.
Enfin un tube en plastique,-reposant ouvert sur le fond de la tranchée, doit
permettre de contrôler une éventuelle accumulation d’eau libre à la base de la
structure grossière.
3. LES MESURES D’HUMIDITE
Les mesures d’humidité en profondeur sont faites avec un humidimètre à
neutrons [4]. Un étalonnage préalable permettant de passer du taux de comptage
des neutrons à l’humidité volumique (Hv = cm3 d’eau/100 cm3 de sol) a été effectué
pour les deux structures (gravier et sol environnant la tranchée).
D existe une zone d’incertitude sur la valeur de l’humidité au voisinage des
interfaces, car la sphère d’influence des neutrons englobe deux matériaux dont
l’humidité est très différente; cette zone d’incertitude des mesures s’étend sur
15 cm du côté de la structure fine humide et sur 25 cm du côté de la structure
sèche.
246 RANÇON
FIG.4. La tranchée après le recouvrement en matériau fin.
Pour tracer les profils hydriques, on fait une mesure tous les 10 cm.
Un pluviomètre enregistreur a été installé à proximité immédiate de la
tranchée.
4. DEROULEMENT DE L’EXPERIENCE
Le dispositif expérimental complet (tranchée, tubes intérieurs et extérieurs,
pluviomètre) a été définitivement implanté à la fin du mois de mai 1973. On a
pris comme origine des temps le 1er juin 1973.
L’étude s’est déroulée en trois phases.
Dans la première phase on a mesuré l’évolution de l’humidité dans la tranchée
et dans le sol environnant sous l’influence des précipitations atmosphériques
pendant un cycle de saisons en effectuant les relevés des profils hydriques à
intervalles de temps réguliers et rapprochés.
IAEA-SM-243/71 247
100 mm pluies
Juin Juillet Août .Septembre. Octobre Novembre Décembre
1974
FIG.5. La pluviométrie pendant la première phase de l ’étude.
La deuxième phase a constitué en des expériences de mise à l’épreuvè de
la barrière capillaire par des irrigations soüs charge de la surface de la tranchée.
La troisième phase s’est effectuée sur une longue durée, 4 ans pendant
lesquels le dispositif a été soumis sans protection à l’action des intempéries, des
mesures régulières de l’humidité dans les diverses structures étant effectuées.
248 RANÇON
Humidité volumique Hv “/•
FIG.6. Evolution des profils hydriques dans le sol environnant la tranchée (phase d ’assèchement).
5. EVOLUTION DE L’HUMIDITE DANS LE DISPOSITIF
ET SES ENVIRONS PENDANT UN CYCLE DE SAISONS
Cette campagne de mesure s’est déroulée de juin 1973 à juin 1974. La
pluviométrie cumulée par semaine est donnée sur la figure 5. Après un été et
un automne plutôt secs malgré de fortes pluies d’orage en octobre, les précipita
tions ont été abondantes et supérieures à la moyenne annuelle de décembre à mai.
5.1. Evolution de l’humidité dans le sol environnant la tranchée
Le tube extérieur à la tranchée (A3) permet de comparer l’évolution de
l’humidité dans le sol environnant la tranchée avec ce qui se passe dans les structures
poreuses constituant le stockage. L’évolution des profils hydriques sur A3 est
montrée sur les figures 6 et 7.
A partir du début des mesures en juin 1973, le sol s’est progressivement
asséché selon le processus normal consécutif à un été sec. Les fortes pluies
IAEA-SM-243/71 249
Humidité volumiquc Hv •/•U 6 8 10 12 U 16 18 20 22 24 26 28
FIG. 7. Evolution des profils hydriques dans le sol environnant la tranchée (phase d ’humidification).
d’octobre ont provoqué une humidification superficielle jusqu’à —40 cm; les
profils sont ensuite restés à peu près stables jusqu’à la mi-décembre. A la suite d’un
hiver et d’un printemps exceptionnellement pluvieux, le sol s’est humidifié pro
gressivement jusqu’à atteindre son état d’humidité maximal en mai 1974.
L’humidité du sol dans son état le plus sec se situe autour de 13% Hv en
profondeur, elle est inférieure à 10% en surface. Dans son état le plus humide,
en mai, le sol est uniformément humidifié aux alentours de 24% selon un profil
continu jusqu’à 3 mètres de profondeur.
5.2. Evolution de l’humidité dans les structures fines de la tranchée
Les tubes traversant la tranchée sont désignés par A l et A2. Les profils
relevés sur Al et A2 étant voisins, on n’a tracé pour la clarté des figures que
le profil A2. D’autre part, comme les profils évoluent progressivement et lente
ment, on en a choisi six parmi les nombreux relevés pour illustrer le comportement
de l’eau dans les structures de la tranchée (fig.8 à 13).
250 RANÇON
150 200 300 <00 500 600 700 900 900 1000 1100 1200 c/9 Humidimttr»
FIG.8. Profils hydriques dans la tranchée et dans le sol environnant, relevés le 19 juin 1973.
La structure fine supérieure, dont l’Hv lors de sa mise en phase se situait
autour de 12%, s’humidifie moins que le sol environnant en raison de la forme
arrondie de la surface qui favorise le ruissellement.
Les premières mesures ont montré une certaine différence entre l’état
hydrique en dessous du fond de la tranchée et celui du sol environnant à la
même cote (fig.8); cela tient au fait que la tranchée est restée ouverte un certain
temps et que le fond a été soumis à l’influence combinée de la pluie et de l’évapora
tion. Au cours de la période d’assèchement l’humidité reste un peu plus importante
sous l’interface inférieure qu’au même niveau du sol environnant (cote —270 cm
sur A2 et —200 sur A3); cet excédent, ne dépassant pas 3,5% Hv (fig.10), se
maintient pendant la phase d’humidification jusqu’à ce que le front d’infiltration
atteigne ce niveau. Cela est dû au fait qu’il y a rupture capillaire au niveau de
l’interface inférieure et que l’eau provenant du bas n’a plus la possibilité de
poursuivre son ascension. Par contre, quand le sol environnant le fond de la
tranchée atteint son taux maximal d’humidification, l’excédent d’humidité sous
l’interface disparaît pour faire place à un léger déficit (fig. 13) car la structure
grossière assure une protection qui s’oppose à la venue des eaux d’infiltration.
IAE A-SM-243/71 251
150 200 300 400 500 600 700 800 900 1000 1100 1200 c Is Humidimttrt
FIG.9. Profils hydriques dans la tranchée et dans le sol environnant, relevés le 9 octobre 1973.
Cette humidité parfois légèrement supérieure sous le fond de la tranchée
ne provient pas d’une accumulation d’eau à partir d’infiltration au travers de la
structure grossière:
— tout d’abord ce léger excédent d’humidité se constate seulement lors des
époques où le sol au même niveau est dans son état de sécheresse maximale
et non pas pendant ses phases d’humidité maximale
— les contrôles réguliers effectués par le tube ouvert reposant sur le fond de la
tranchée n’ont jamais décelé la présence d’eau libre pas plus que de traces
de boues
— enfin, ni les fortes pluies, ni l’expérience de rejet de 400 mm d’eau à la surface
de la tranchée n’ont été suivies d’une augmentation du taux d’humidité au
niveau de l’interface inférieure.
5.3. La structure sèche
Pendant cette période d’un an les variations du taux de comptage des profils
neutroniques ne révèlent pas d’humidification sensible de la structure grossière.
252 RANÇON
150 200 300 «00 500 600 700 S00 900 1000 1100 1200 c/s Humidimitr»
FIG. 10. Profils hydriques dans la tranchée et dans le sol environnant, relevés le 17 janvier 1974.
Le taux de comptage le long des profils neutroniques ou à une cote donnée
en fonction du temps fluctue entre 148 et 164 coups-s-1 (tableau I). Des séries
de mesures successives sur un point donné ont montré, d’autre part, que le taux
de comptage variait de ±2 coups-s-1 autour de la valeur moyenne pour un temps
de comptage usuel de 50 secondes.
En théorie, 151 ± 2 coups-s-1 correspondent à 0% d’humidité dans le gravier.
En fait le taux de comptage du point zéro dans la structure grossière est sujet à
plusieurs types de fluctuations: fluctuations statistiques, fluctuations dans le
temps dues à l’électronique (charge, température) et fluctuations dans l’espace
dues à la structure (variations de tassement, de classement des grains).
En sachant que 1% d’humidité volumique donnerait un taux de comptage
de 190 coups-s_1 et d’autre part qu’en dehors de toutes fluctuations un taux de
160 coups-s-1 correspondrait à 0,2% Hv, valeur trop faible pour être significative,
on peut considérer que les variations autour de 0% théorique ne sont pas signi
ficatives d’une humidification de la tranchée.
Ainsi, pendant le cycle saisonnier au cours duquel la tranchée a subi l’influence
des intempéries et malgré un hiver particulièrement pluvieux, on n’a pas relevé
d’humidification sensible de la structure grossière, que l’on peut qualifier de
structure sèche.
IAEA-SM-243/71 253
« С 200 300 too 500 600 700 800 SOO 1000 1100 1200 c/s Humidimitre
FIG.11. Profils hydriques dans la tranchée et dans le sol environnant, relevés le 4 février 1974.
Toutefois, on a jugé nécessaire, pour que l’expérience soit plus probante,
de soumettre la tranchée à un arrosage artificiel dans des conditions beaucoup
plus sévères que celles dues aux seules pluies.
6. EXPERIENCE DE REJET CONTINU DE GRANDS VOLUMES
D’EAU A LA SURFACE DE LA TRANCHEE
6.1. Conduite de l’expérience
On a creusé une cuvette circulaire centrée autour du tube A2 sur une pro
fondeur de 5 cm et une surface de 2000 cm2. L’eau est rejetée dans la cuvette par
fractions de 10 litres, ce qui correspond à une hauteur d’eau de 50 mm. Après
infiltration de chaque rejet de 50 mm, on mesure le profil hydrique, puis on
procède au rejet de la fraction suivante. La vitesse d’infiltration, plus rapide au
début, se stabilise autour de 75 m m i '1.
254 RANÇON
150 200 300 100 500 600 700 800 900 1000 1100 1200 c/s Humidimttre
FIG.12. Profils hydriques dans la tranchée et dans le sol environnant, relevés le 5 mars 1974.
On arrête le rejet après infiltration de 400 mm et on observe l’évolution des
profils hydriques pendant la redistribution de l’eau. (Le terme infiltration désigne
les mouvements d’eau dans le sol pendant la durée d’un apport extérieur, la
redistribution les mouvements d’eau en l’absence d’apport extérieur.)
6.2. Evolution des profils hydriques au-dessus de l’interface supérieure
6.2.1. Pendan 11 ’infil tra tion
Avant d’atteindre l’interface de séparation des deux structures, le front
d’humidité progresse régulièrement dans la structure fine supérieure. On voit sur
la figure 14 les profils hydriques après 100, 200 et 300 mm infiltrés; les fronts
d’humidité sont bien parallèles.
Après infiltration de 350 mm, le front d’humidité atteint l’interface de
séparation des deux structures, après infiltration de 400 mm le front d’humidité
est confondu avec le précédent, la progression vers le bas est arrêtée (fig.l4).
IAEA-SM-243/71 255
150 200 300 too 500 600 700 BOO 900 1000 1100 1200 c / s Humidimttrt
FIG.13. Profils hydriques dans la tranchée et dans le sol environnant, relevés le 14 mai 1974.
6.2.2. Pendant la redistribution (fig. 15)
Les rejets sont arrêtés après infiltration de 400 mm afin d’observer la redistri
bution de l’eau dans le sol.
Pendant la phase d’infiltration de 350 à 400 mm et à la fin de l’infiltration,
le sol est fortement humidifié autour de 33% Hv; les eaux liées et gravitaires,
uniformément réparties entre la surface et l’interface, sont en équilibre avec l’eau
en charge dans la cuvette.
Après la fin de l’infiltration dans une redistribution normale, il y aurait
continuation de l’avancement du front avec une diminution consécutive de l’eau
le long du profil. Dans le cas présent (fig. 15) les profils mesurés 4 et 20 heures
après l’infiltration montrent qu’il y a redistribution de l’eau, dont la teneur
diminue rapidement tout le long du profil jusqu’à l’interface, mais que cette eau
ne se retrouve pas en dessous du front d’humidité mesuré à la fin de l’infiltration.
D n’y a pas de progression du front; l’eau a été drainée latéralement le long de
l’interface qui a la forme d’un toit.
TABLEAU I. TAUX DE COMPTAGE DES PROFILS NEUTRONIQUES DANS LA STRUCTURE GROSSIERE
PENDANT UN CYCLE DE SAISONS (coups s-1)
Cote
(-cm)
1973 1974
Moyenne
19 juin 21 août 14 sept. 9 oct. 5 déc. 17 janv. 4 févr. 5 mars 19 avril 14 mai
100 156 163 162 155 151 158 163 152 151 152 156,3
110 152 166 160 152 141 147 159 145 144 145 151,1
120 152 163 158 150 146 145 150 143 143 145 149,4
130 153 153 160 152 149 148 148 142 144 147 149,6
140 157 158 161 151 150 152 152 142 149 146 151,8
150 159 158 160 155 148 151 154 147 151 152 153,8
160 162 160 158 157 157 155 156 149 151 156 155,9
170 161 159 162 157 160 157 162 151 157 156 158,2
180 164 161 160 162 164 159 160 154 155 157 159,6
190 160 155 159 166 156 160 162 154 152 154 156,8
200 161 157 158 154 157 155 161 155 154 153 156,5
Moy. 157,9 159,4 159,8 154,5 152,6 153,4 156,9 148,6 150,1 151,2 154
256 RAN
ÇON
IAEA-SM-243/71 257
Humidité voíumíque •/•I terre )
FIG. 14. Evolution des profils hydriques dans la structure fine pendant l ’infiltration.
Cette mise à l’épreuve montre bien l’efficacité de la barrière capillaire si
on considère qu’un arrosage de 400 mm, éliminant tout effet de ruissellement le
long du dôme de la tranchée et effectué en quelques heures, constitue la moitié
des précipitations annuelles.
7. UTILISATION DE L’IODE-131 COMME DETECTEUR DE FUITES
Le traceur radioactif a été utilisé non pas pour suivre l’avancement du front
d’humidité mais pour détecter les voies d’eau éventuelles susceptibles de se pro
duire au travers de l’interface sous l’effet d’une trop forte charge. Ces fuites,
comme on l’a vu sur les maquettes, se produisent sous forme de filets d’eau, elles
risquent ainsi de ne pas être détectées par la sonde à neutrons, qui explore une
FIG. 15. Redistribution des 400 mm d ’eau infiltrés.
Ю 50 Ю0 C.S-1
FIG. 16. Progression du traceur radioactif.
IAEA-SM-243/71 259
I ;150 200 300 400 500 600 TOO-' 800 900 1000 1100 1200 1300
FIG.17. Profils hydriques dans la tranchée et dans le sol environnant, relevés le 13 août 1976.
sphère; par contre elles pourraient être décelées par un radiamètre grâce au
rayonnement gamma du radioélément.
L’iode-131 sans entraîneur n’est pas un bon marqueur de l’eau dans le sol
car il est retenu par certains minéraux [4]. C’est pourquoi on y a ajouté de
l’entraîneur Nal (0,01N). Malgré cela le front d’activité se propage un peu plus
lentement que le front d’humidité (décalage de 15 cm).
On voit sur la figure 16 que les fronts d’activité se superposent au niveau de
l’interface; de même pendant la redistribution, il y a diminution de l’activité le
long du profil, mais pas de progression du front. La sonde détecte évidemment
une certaine activité décroissante au-dessous de l’interface, due à la pénétration
du rayonnement; au-dessous de cette zone d’incertitude aucune activité supérieure
au bruit de fond n’a été décelée dans la structure grossière.
260 RANÇON
! » 200 300 (00 500 600 TOO 800 900 Ю00 1100 1200 1300
FIG. 18. Profils hydriques dans la tranchée et dans le sol environnant, relevés le 28 novembre 1977.
8. CONTROLES SUR UNE LONGUE DUREE (1974-79)
Après l’expérience de mise à l’épreuve par irrigation, on a attendu que la
structure fine retrouve son équilibre hydrique naturel et on a repris les mesures
à des intervalles plus espacés de 1974 à 1979 pour faire porter les contrôles sur
une longue durée.
Les résultats concernant les profils hydriques sont indiqués sur les figures 17
à 20 et sur le tableau II. On n’a pas décelé d’humidification sensible de la
structure grossière d’après ces profils. Les contrôles effectués dans le tube ouvert
n’ont pas décelé la présence d’eau ou de boue dans le fond de la tranchée.
Il convient de tenir compte des changements survenus sur la sonde au cours
de ces années. Il est usuel avant chaque profil de faire une mesure de référence
dans l’étui de paraffine qui sert de protection contre les neutrons quand la source
se trouve dans l’appareil. Cette mesure de référence, sujette à des fluctuations,
IAEA-SM-243/71 261
ISO 200 300 400 500 600 700 800 900 1000 1Ю0 ООО ООО
FIG.19. Profils hydriques dans la tranchée et dans le sol environnant, relevés le 21 juin 1978.
a varié de façon irrégulière de 790 à 800 coups-s-1 pendant la première année
d’observation. Ces variations se sont accentuées à partir de 1976, peut-être en raison
du vieillissement de l’appareil. On voit par exemple sur le tableau II que le
21 juin 1978 le profil hydrique avec une mesure étui de 738 coups's-1 est nette
ment en dessous du % Hv théorique correspondant à 151 coups-s-1 avec
800 coups's-1 de mesure étui. Dans ces cas il faut donc effectuer une correction
pour réévaluer le 0% Hv en fonction des variations des performances de l’appareil
de mesure.
9. LIMITES ET AVANTAGES DE CETTE METHODE DE STOCKAGE
9.1. Limites et précautions indispensables
L’efficacité de la barrière capillaire pour la protection contre l’infiltration
a été prouvée sur un dispositif de dimensions réduites pour lequel la mise en place
et le choix des matériaux constituant les structures ont été soignés.
262 RANÇON
ISO 200 300 400 500 600 700 800 900 1000 1100 1200 1300
FIG.20. Profils hydriques dans la tranchée et dans le sol environnant, relevés le 13 mars 1979.
Dans le cas général, il y aura essentiellement deux impératifs à respecter:
— La structure fine supérieure doit être mise en place de telle sorte qu’il
ne puisse pas y avoir de chemins préférentiels entraînant des circulations d’eau
libre à la suite de fissures ou hétérogénéités diverses. D faut utiliser un matériau
dont les grains sont obligatoirement inférieurs à 0,25 mm [ 1 ], mais ce matériau
ne doit pas être trop argileux ni trop consolidé pour ne pas se trouver, lors de sa
mise en place, sous forme agglomérée en mottes ou en gros grains et constituer
ainsi un milieu grumeleux de grande perméabilité; c’est pourquoi la terre de'
déblai n’est pas toujours propice à la constitution de la structure fine à moins
d’avoir été au préalable pulvérisée, opération difficile sur de grandes quantités.
Un bon matériau serait un sable argileux avec une proportion suffisante de sable
pour qu’il reste pulvérulent sans devenir plastique à l’humidification, ni former
des craquelures à la dessiccation.
— L’interface doit être nette. D faut éviter le mélange des deux structures
si les grains du matériau fin sont plus petits que les pores du,matériau grossier,
ce qui, en rendant la surface de séparation plus diffuse, nuirait à l’efficacité de
l’écran capillaire. On peut choisir un matériau grossier dont la granulométrie
IAEA-SM-243/71 263
TABLEAU II. TAUX DE COMPTAGE DES PROFILS NEUTRONIQUES
DANS LA STRUCTURE GROSSIERE DE 1975 A 1979 (coups-s-1)
Cote
(-cm)
20 juin
1975
13 août
1976
28 nov.
1977
21 juin
1978
13 mars
1979
100 155 153 159 146 148
110 157 151 154 140 138
120 159 156 154 140 140
130 154 152 154 136 140
140 155 151 157 140 148
150 156 153 157 140 146
160 157 156 163 146 152
170 157 158 164 146 148
180 157 160 163 140 150
190 152 152 163 140 150
200 157 156 166 136 146
Moyenne 156 154 159 141 148
Réf. étui 807 769 852 738 767
Réévaluation
0% HV a
152 148 161 139 145
a Pour 151 coups"s 1, référence étui de 800 coups-s l .
ne permet pas cette interpénétration comme cela a été fait sur les modèles réduits
[i]; mais ce choix peut s’avérer difficile dans la pratique. On peut aussi pallier
à ce risque en consolidant la terre fine au niveau de l’interface ou en interposant
une membrane poreuse, telle que la laine de verre dans la tranchée décrite ici ou
diverses toiles poreuses résistantes utilisées dans des travaux de génie civil.
9.2. Avantages
Le matériau grossier, outre son rôle de support de la barrière capillaire,
présente un autre avantage: ses grains sont assez gros pour ne pas s’agglomérer
après humectation; il garde aussi une fluidité qui, lors de la mise en place,
permet d’entourer les déchets de façon homogène en empêchant la formation de
cavités qui risquent d’engendrer des affaissements. On obtient ainsi avec ce type
de matériau un ensemble plus compact qu’avec une terre de déblai argileuse.
264 RANÇON
Le matériau grossier, support de la structure sèche, est peu coûteux. Sa
mise en place en masse homogène est aisée et il n’y a pas de difficultés à constituer
des interfaces franches avec le fond et les flancs de la tranchée en général creusée
dans un sol consolidé.
Enfin si pour diverses raisons l’installation devait être démantelée et les
déchets récupérés, il serait beaucoup plus facile de réaliser l’enlèvement d’un
matériau granuleux, sec, peu générateur de poussières et non adhérent aux déchets
que de tout autre matériau pulvérulent.
10. CONCLUSION
La tranchée expérimentale telle qu’elle a été construite, avec ses dimensions
assez réduites, a bien assumé son rôle qui était de maintenir dans le sol une structure
sèche pendant une longue durée et malgré des conditions rigoureuses d’arrosage
superficiel.
Cette technique peut donc s’adapter à l’amélioration de la sûreté des stockages
de déchets par enfouissement. U serait pourtant souhaitable de la tester sur un
stockage réel dans lequel seraient disposés des appareillages nécessaires à des
mesures précises d’humidité.
On a objecté qu’une protection indéfinie n’est pas garantie; on peut répondre
qu’aucun des dispositifs de ce type ne peut prétendre à long terme à une absolue
efficacité, mais que le procédé des barrières capillaires, simple, autant et même
plus robuste que les dispositifs classiques, tout en restant peu coûteux, apporte
en plus une amélioration notable à la sûreté des stockages par enfouissement de
déchets de faible ou moyenne radioactivité grâce à la formation de structures sèches.
REFERENCES
[1] RANÇON, D., Structures sèches et barrières capillaires en milieux poreux, Application au
stockage dans le sol, Rapport CEA-R-4310 (1972).
[2] BEAUGELIN, F., BEYER, G., RANÇON, D., Structures sèches et barrières capillaires en
milieu poreux, Film, Bureau de documentation, CEN de Cadarache, St-Paul-lez-Durance.
[3] BARBREAU, A., MARCAILLOU, J., MERY, J., PINTO, D„ RANÇON, D„ «Evolution
de la gestion des déchets de basse activité et d’activité intermédiaire au Centre de
Cadarache », Management of Low- and Intermediate-Level Radioactive Wastes (C.R.
Coll. Aix-en-Provence, 1970), AIEA, Vienne (1970) 347.
[4] RANÇON, D., «Utilisation simultanée d’humidimètres à neutrons et d’émetteurs gamma
pour les mesures in situ des mouvements de l’eau et des ions dans les sols non saturés»,
Isotope and Radiation Techniques in Soil Physics and Irrigation Studies 1973 (C.R. Coll.
Vienne, 1973), AIEA, Vienne (1974) 225.
IAEA-SM-243/71 265
DISCUSSION
L.J. ANDERSEN: What is the optimal ratio between the grain sizes of the
fine and the coarse material, and what is the optimal grain size of the fine
material? The problem seems to be to establish a hydraulic or capillary conduc
tivity within the fine material which is sufficient to transport the recharged water
under the actual conditions. Have you studied the water flux in the vapour phase
of the coarse material?
D.L. RANÇON: The coarse material is intended only to provide support for
the fine material and is not involved in the transfer of water. The capillary barrier
consists entirely of the fine material whose grains must be smaller than 0.2 mm
in size (see Ref. [ 1 ]). However, from a purely physical standpoint, there is an
optimal ratio between the two grain sizes for the formation of an ideal system;
the grains of the fine material should not penetrate into the pores of the coarse
material, as was the case in the reduced-scale models (Ref. [ 1 ]) in which the grain
size of the fine sand was between 0.1 and 0.2 mm and that of the coarse sand
between 0.4 and 0.8 mm.
We have not performed any measurements of transfer in the vapour phase.
However, the overall measurements did not indicate any detectable condensations.
J. HOWIESON: What happens after the coarse material has become saturated?
Does it then provide a drain to the system?
D.L. RANÇON: The purpose of the experiment is to prevent saturation of
the coarse material. However, if such an accident occurred, as the fine medium
regained sufficient dryness to ensure suitable suction, the water would be sucked
out of the coarse medium by the fine medium.
D.F. DIXON: As we have done similar work at the University of Waterloo
in Canada, I should like to ask if you have established static hydraulic break
through pressures for the systems you are employing.
D.L. RANÇON: We have not performed quantitative measurements of the
hydraulic pressure. All we have done is to measure the moisture rate in order to
evaluate the presence of various forms of water (capillary water, gravity water)
in the medium.
DISPOSAL IN DEEP GEOLOGICAL FORMATIONS:
EVAPO RITES
(Session IV)
Chairman
K. KÜHN
Federal Republic of Germany
IAEA-SM-243/38
CHARACTERIZATION OF A SITE
IN BEDDED SALT FOR ISOLATION OF
RADIOACTIVE WASTES
L.R. HILL
Sandia Laboratories,
Albuquerque, New Mexico,
United States of America
Abstract
CHARACTERIZATION OF A SITE IN BEDDED SALT FOR ISOLATION OF
RADIOACTIVE WASTES.
The first radioactive waste repository in the USA is scheduled to begin operation in the
1980s. Plans are to locate this facility, the Waste Isolation Pilot Plant (WIPP) in southeast
New Mexico as a repository for transuranic (TRU) waste from past and current United States
defence programmes. In addition, the WIPP would provide a research facility to examine, on
a large scale, the interactions between bedded salt and high-level radioactive waste. The United
States Department of Energy has recommended that WIPP also be used to demonstrate surface
and subsurface methods of handling, storing and disposing of up to 1000 canisters of spent
reactor fuel; however, a decision to implement this recommendation has not been made. The
proposed underground storage facilities aie to be placed near the middle of a 3600-foot-thick
sequence of relatively pure evaporite strata containing primarily rock salt. Within this sequence,
the formation richest in rock salt is nearly 2000 feet thick and contains the two proposed
underground storage levels at depths near 2100 feet and 2650 feet. In the past four years,
many techniques have been used to characterize the WIPP site. Geophysical surveys include
about 140 line miles of new seismic reflection data and over 9000 resistivity measurements.
More than 50 boreholes have been drilled. Hydrologie studies of the proposed site and adjacent
area have been directed toward a quantitative evaluation of the salt-dissolution process, the
hydrogeologic parameters affecting groundwater movement, and the major elements of surface
and groundwater quality as related to water-resource use. Potash salt and natural gas are two
potential resources of economic significance under the WIPP site. Potassium salts occur in a
variety of mineral types, but only sylvite and langbeinite are currently economic. It is believed
that a langbeinite deposit located in the northeast quadrant of the WIPP site could be profitably
mined using today’s technology. Several deposits of sylvite are present in the WIPP site, but
none would be developed under today’s economic conditions. Natural gas is estimated to be
economically recoverable from the site area; however, its value is insignificant compared to
total US reserves. Laboratory and in situ experimental programmes are to be performed
before the WIPP facility is operational and will address such issues as brine migration, corrosion,
hole plugging and structural deformations. These programmes are aimed at refining and
supplementing information gathered to increase confidence in the use of this site. The
information gathered will also support repository design and long-term safety assessment.
269
270 HILL
1. INTRODUCTION
This paper focuses on the Waste Isolation Pilot Plant (WIPP) being planned in the United States, its site selection criteria, the status of site selection studies, results of these studies to date and the programs that remain to be completed. The Geological Characterization Report Waste Isolation Pilot Plant (WIPP), Southeastern New Mexico, SAND78-1596, Powers et al., December, 1978, [1] is the source of much of the material presented here.
2. PURPOSE OF WIPP
The purpose of the WIPP is to demonstrate the technology for the disposal of the transuranic (TRU) waste resulting from the past and current US defense programs. It is planned to convert the WIPP into a repository after successful demonstration of this technology and assessment of safety of a repository in southeastern New Mexico. In addition, the WIPP is to provide a research facility to examine, on a large scale, the interactions between bedded salt and high-level radioactive waste. A Department of Energy (DOE) Task Force [2 ] has recommended that WIPP also be used to demonstrate surface and subsurface methods of handling, storing and disposing of up to 1 000 canisters of spent reactor fuel. A decision on implementing this recommendation has not been made at this time.
If this site is accepted by the DOE, the schedule calls for the initiation of facility construction in early 1981; completion is to be late 1985, and the first waste to be accepted in 1986. The TRU waste would be readily retrievable for a five to ten year period of initial operation. All HLW for experiments would be retrieved upon completion of the experiments. The conceptual design of WIPP facilities is complete [3]. Detailed facilities design is new being performed.
3. SITE SELECTION
3.1 Historical Development
Interest in disposal of radioactive waste in geologic media may be traced back to a 1957 committee report by the US National Academy of sciences - National Research Council [4] that recommended guidelines for permanent disposal of radioactive waste in geological media. The recommendations may be placed in two categories: burial in bedded salt deposits or indeep sedimentary basins (perhaps 4 000 - 5 000 m deep).
IAEA-SM-243/38 271
Е Ш STUDY АЙЕА
• DOMES OR CIRCULAR STRUCTURES
- - POSTULATED FAULTS
- -« i. CAPITAN REEF
^ ^ D IS S O L U T IO N FRONT
____ POTASH ENCLAVE
------ PAVED ROAO
Index Map
NewMexico
О 10 20 Miles
0 10 20 30 K ilom eters
FIG .l. Map of southeast New Mexico.
Salt became the leading candidate as a disposal medium, and from 1957 until the 1970's most disposal studies in the USA concentrated on bedded salt. In the mid-1960's, Oak Ridge National Laboratories (ORNL) conducted a successful in situ experimental program called "Project Salt Vault" in a salt mine near Lyons, Kansas. A subsequent plan to establish a federal repository near Lyons was withdrawn due to both technical and political objections.
Subsequent evaluation of salt basins in the United States by ORNL and the US Geological Survey led in 1974 to field investigations of Permian salt deposits of the Delaware Basin in southeastern New Mexico to determine if the geologic setting was adequate for a radioactive waste repository, Figure 1. Permian evaporite deposits consist of the Castile Formation which is inter bedded halite and anhydrite, the Salado Formation, which consists principally of halite, and the Rustler Formation, which is mostly anhydrite but contains halite, dolomite, and siltstone, Figure 2.
In 1975, the Atomic Energy Commission, now Department of Energy (DOE), assigned responsibility for site evaluation, environmental assessment and conceptual design of facilities for this project in New Mexico to Sandia Laboratories of Albuquerque, New Mexico. The project in New Mexico is now known as the Waste Isolation Pilot Plant (WIPP).
272 HILL
FIG.2. Geologic section through the WIPP area.
3.2 Site Selection Criteria
Site selection criteria were developed to meet the desired goal of complete isolation of radioactive waste with negligible consequence in the event of containment failure for the duration of time in which the radioactivity could constitute a potential hazard to the biosphere or humans in general. At first, preliminary site selection criteria were general in nature. (Most areas are not sufficiently known to allow application of precise criteria.) After Sandia Laboratories determined in 1975 that the first preliminary study area was geologically unsuitable, site selection criteria and the factors which address the criteria were refined specifically for,and applied to,the Delaware Basin in New Mexico to define the present study area.
The site selection criteria and factors considered for WIPP are summarized in the following paragraphs. It should be
IAEA-SM-243/38 273
emphasized that these factors are desirable goals, but in and of themselves may not be necessary to assure safety of the repository. The consequences of real processes and postulated failure scenarios will be calculated and used in the assessment of site acceptability. The total system of multiple barriers must be considered in the evaluation of site acceptability and not simply isolated criteria. This safety assessment is presented in the draft environmental impact statement for WIPP [5].
3.2.1 Geology Criterion: The geology of the site will be suchthat the repository will not be breached by natural phenomena while the waste poses a significant hazard to man. The geology must also permit safe operation of the WIPP.
Geology Factors:
Topography - as related to access for transportation, effects on inducing salt flow, determining surface water flow and chances of future inundation.
Depth - as related to mine stability and protection frcm erosion and consequences of surficial phenomena.
Thickness - as related to buffering thermal and mechanical effects.
Lateral Extent - as related to the distance to structural or salt dissolution boundaries.
Lithology - as related to the purity of the salt beds and the brine content of the salt.
Stratigraphy - as related to the continuity of beds, character of inter-bedding and nature of beds over- and underlying the salt.
Structure - as related to relatively flat bedding (<3°) for operational purposes. Steep anticlines and major faults are to be avoided.
Erosion - as related to features which would tend to localize an^/or accelerate erosion.
Dissolution - as related to a breach of the repository while the wastes represent a significant hazard to man.
Subsidence - as related to adverse affects on the repository beds or undue acceleration of the rate of
274 HILL
d i s s o l u t i o n t o t h e j e o p a r d y o f l o n g - te r m i n t e g r i t y o f th e r e p o s i t o r y .
3 . 2 . 2 H y d ro lo g y C r i t e r i o n : The h y d r o lo g y o f t h e s i t e m u s tp r o v id e h ig h c o n f id e n c e t h a t n a t u r a l d i s s o l u t i o n w i l l n o t b r e a c h t h e s i t e w h i le t h e w a s te p o s e s a s i g n i f i c a n t h a z a r d t o m an . A c c i d e n t a l p e n e t r a t i o n s s h o u ld n o t r e s u l t i n u n d u e h a z a r d s t o m a n k in d .
H y d ro lo g y F a c t o r s :
S u r f a c e W ate r - a s r e l a t e d t o p r e s e n t an d f u t u r e r u n o f f p a t t e r n s , f l o o d in g p o t e n t i a l , e t c .
A q u i f e r s - a s r e l a t e d t o the> o v e r - an d u n d e r ly i n g a q u i f e r s w h ic h f o r WIPP r e p r e s e n t a s e c o n d a ry b a r r i e r i f t h e s a l t i s b r e a c h e d .
H y d r o lo g ie T r a n s p o r t - a s r e l a t e d t o a s e c o n d a r y f a c t o r w h ic h , f o r t h e W IPP, m u s t be e v a l u a t e d t o a l l c w q u a n t i t a t i v e c a l c u l a t i o n s o f th e c o n s e q u e n c e s o f v a r i o u s f a i l u r e s c e n a r i o s .
C l i m a t i c F l u c t u a t i o n s - a s r e l a t e d t o p o s s i b l e p l u v i a l c y c l e s .
M an-m ade P e n e t r a t i o n s - a s r e l a t e d t o t h e e f f e c t o f d r i l l h o le s a n d m in in g o p e r a t i o n s o n t h e s i t e .
3 . 2 .3 T e c to n ic s t a b i l i t y C r i t e r i o n : N a t u r a l t e c t o n i cp r o c e s s e s m u s t n o t r e s u l t i n a b r e a c h o f t h e s i t e w h i le t h e w a s t e s r e p r e s e n t a s i g n i f i c a n t h a z a r d t o man an d s h o u ld n o t r e q u i r e e x t r a ñ e p r e c a u t i o n s d u r in g t h e o p e r a t i o n a l p e r i o d o f t h e r e p o s i t o r y .
T e c to n ic S t a b i l i t y F a c t o r s :
S e is m ic A c t i v i t y - a s r e l a t e d t o t h e f r e q u e n c y an d m a g n i tu d e o f s e i s m ic a c t i v i t y w h ic h a f f e c t s f a c i l i t y d e s ig n an d s a f e t y o f o p e r a t i o n .
F a u l t i n g / F r a c t u r i n g - a s r e l a t e d t o o p en f a u l t s , f r a c t u r e s o r j o i n t s . W h ile n o t e x p e c t e d i n s a l t , t h e m ore b r i t t l e u n i t s w i t h i n and s u r r o u n d in g th e s a l t may s u p p o r t su c h f e a t u r e s w h ich c o u ld e n h a n c e d i s s o l u t i o n and h y d r o l o g i e t r a n s p o r t . M a jo r f a u l t s an d p ro n o u n c e d l i n e a r s t r u c t u r a l t r e n d s s h o u ld be a v o id e d .
S a l t F l c w / A n t i c l i n e s - a s r e l a t e d t o m a jo r d e f o r m a t io n o f s a l t b e d s by f lo w w h ic h c a n f r a c t u r e b r i t t l e r o c k an d c r e a t e
IAEA-SM-243/38 275
p o r o s i t y f o r b r i n e a c c u m u la t io n s . M a jo r a n t i c l i n e s r e s u l t i n g f r a n s a l t f lo w s h o u ld be a v o id e d o r e v a l u a t e d to c h e c k on b r i n e p r e s e n c e an d a n h y d r i t e f r a c t u r i n g .
D ia p i r i s m - a s r e l a t e d t o an e x t r e m e r e s u l t o f s a l t f l e w .
R e g io n a l S t a b i l i t y - a s r e l a t e d t o a r e a s o f p r o n o u n c e d r e g i o n a l u p l i f t o r s u b s i d e n c e s i n c e s u c h b e h a v io r m akes a n t i c i p a t i o n o f f u t u r e d i s s o l u t i o n , e r o s i o n an d s a l t f lo w m ore u n c e r t a i n .
I g n e o u s A c t i v i t y - a s r e l a t e d t o a r e a s o f a c t i v e o r r e c e n t v o lc a n is m o f ig n e o u s i n t r u s i o n s h o u ld b e a v o id e d t o m in im iz e t h e s e h a z a r d s t o t h e r e p o s i t o r y .
G e o th e rm a l G r a d i e n t - a s r e l a t e d t o a b n o r m a l ly h ig h g e o th e r m a l g r a d i e n t s . H igh g r a d i e n t s may be i n d i c a t i v e o f r e c e n t ig n e o u s o r t e c t o n i c a c t i v i t y .
3 . 2 . 4 P h y s ic a l - C h e m ic a l C o m p a t i b i l i t y C r i t e r i o n : Ther e p o s i t o r y m edium m u s t n o t i n t e r a c t w i th t h e w a s te i n w ays w h ich c r e a t e u n a c c e p ta b l e o p e r a t i o n a l o r lo n g - te r m h a z a r d s .
P h y s ic a l - C h e m ic a l F a c t o r s :
F l u i d C o n te n t - a s r e l a t e d t o t h e f l u i d c o n t e n t o f th e r e p o s i t o r y bed c o n t a i n i n g h i g h - l e v e l w a s t e .
T h e rm a l p r o p e r t i e s - a s r e l a t e d t o m a jo r n a t u r a l th e r m a l b a r r i e r s w h ic h may c a u s e u n d e s i r a b l e t e m p e r a t u r e r i s e s .
M e c h a n ic a l P r o p e r t i e s - a s r e l a t e d t o th e s a f e e x c a v a t io n o f o p e n in g s e v e n w h i l e t h e r m a l l y l o a d e d .
C h e m ic a l p r o p e r t i e s / M i n e r a l o g y - a s r e l a t e d t o b e d s o f u n u s u a l c o m p o s i t io n a n d /o r c o n t a i n i n g m in e r a l s w i th bou n d w a te r n e a r th e w a s te h o r i z o n .
R a d i a t i o n E f f e c t s - a s r e l a t e d t o p o s t u l a t e d d e l e t e r i o u s e f f e c t s .
P e r m e a b i l i t y - a s r e l a t e d t o p e r m e a b i l i t y o f s a l t an d th e i n t e r - b e d s an d s u r r o u n d i n g m e d ia . ( S a l t p e r m e a b i l i t y t o g a s e s may be im p o r ta n t i n e s t a b l i s h i n g w a s t e - a c c e p t a n c e c r i t e r i a . )
N u c l i d e M o b i l i t y - a s r e l a t e d t o a s e c o n d a r y f a c t o r i n s i t i n g , s i n c e c o n f in e m e n t by t h e s a l t an d i s o l a t i o n fro m w a te r i s t h e b a s i c i s o l a t i o n p r e m is e .
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3 . 2 .5 E c o n o m ic /S o c i a l C o m p a t i b i l i t y C r i t e r i o n : The s i t es h o u ld b e o p e r a b l e a t r e a s o n a b l e e c o n o m ic c o s t an d s h o u ld n o t c r e a t e u n a c c e p ta b le im p a c t on n a t u r a l r e s o u r c e s o r th e b i o l o g i c a l / s o c i o l o g i c a l e n v i r o r n e n t .
E c o n o m ic /S o c ia l F a c t o r s :
N a t u r a l R e s o u rc e s - a s r e l a t e d t o m in im iz in g c o n f l i c t o f t h e r e p o s i t o r y w i th a c t u a l o r p o t e n t i a l r e s o u r c e s .
M an-m ade P e n e t r a t i o n s - a s r e l a t e d t o b o r e h o le s o r s h a f t s w h ich p e n e t r a t e t h r o u g h t h e s a l t i n t o u n d e r ly i n g a q u i f e r s w i t h i n on e m i l e o f th e r e p o s i t o r y .
T r a n s p o r t a t i o n - a s r e l a t e d t o e a s e o f s y s te m d e v e lo p m e n t .
A c c e s s i b i l i t y - a s r e l a t e d t o a c c e s s i b i l i t y f o r t r a n s p o r t a t i o n and u t i l i t i e s .
L and J u r i s d i c t i o n - a s r e l a t e d t o s i t i n g o n f e d e r a l l y c o n t r o l l e d l a n d t o t h e e x t e n t p o s s i b l e .
P o p u la t i o n D e n s i ty - a s r e l a t e d t o p r o x im i t y t o p o p u l a t i o n c e n t e r s an d r u r a l h a b i t a t s .
E c o l o g i c a l E f f e c t s - a s r e l a t e d t o m a jo r im p a c ts on e c o lo g y d u e t o c o n s t r u c t i o n an d o p e r a t i o n .
S o c i o l o g i c a l I m p a c ts - a s r e l a t e d t o d e m o g ra p h ic an d e c o n o m ic e f f e c t s .
3 .3 P r e l i m i n a r y S i t e S e l e c t i o n
S in c e o n ly tw o a l t e r n a t i v e s i t e s i n t h e New M ex ico p a r t o f t h e D e la w a re B a s in w i th s t o o d t h e s e t o f s i t i n g c r i t e r i a , t h e p r e l i m i n a r y s e l e c t i o n o f a p r e f e r r e d s i t e w as f a i r l y s t r a i g h t f o r w a r d . A l t e r n a t e I , new known a s t h e L os M edaños s i t e , a p p e a r e d t o be t h e p r e f e r r e d s i t e . A l t e r n a t e I I was c o n s i d e r e d l e s s d e s i r a b l e b e c a u s e i t i s r e s t r i c t e d i n s i z e , t h e a c c e p t a b l e s a l t z o n e s a r e d e e p e r , i t i s c l o s e r t o m a jo r o i l f i e l d s and t o t h e a r e a s o f d e e p d i s s o l u t i o n i n g t o t h e s o u t h , a n d th e h i g h - p u r i t y s a l t l y i n g b e tw e e n t h e Cowden a n h y d r i t e ( i n t h e lo w e r S a la d o F o rm a tio n ) an d t h e C a s t i l e i s th o u g h t t o be a b s e n t . The t o p o f th e S a la d o i s a b o u t 800 f e e t d e e p a t L os M edaños ( s e e F ig u r e 2) v e r s u s 1 500 f e e t a t A l t e r n a t e I I .
S a n d ia L a b o r a t o r i e s s e l e c t e d th e L os M edaños a l t e r n a t e a s t h e b e s t c a n d id a t e a r e a i n e a r l y D ec em b e r, 1 9 7 5 . G e o lo g i c a l c h a r a c t e r i z a t i o n a c t i v i t i e s w ere th e n e x p a n d e d t o f o c u s on
IAEA-SM-243/38 277
obtaining subsurface data at the Los Medaños site. A descriptive summary of these programs follows.
4 . STATUS OF SITE SELECTION STUDIES
The g e o l o g i c s t u d i e s f o r t h e WIPP f a l l n a t u r a l l y i n t o t h r e e d i f f e r e n t p h a s e s : p r e l i m i n a r y s i t e s e l e c t i o n a c t i v i t i e s , g e o l o g i c a l c h a r a c t e r i z a t i o n , an d s t u d i e s o f lo n g - r a n g e g e o l o g i c p r o c e s s e s a f f e c t i n g a r e p o s i t o r y . P r e l i m i n a r y s i t e s e l e c t i o n a c t i v i t i e s a r e c o m p le te now; t h e s e c o n s i s t e d p r i m a r i l y o f n a t i o n a l an d r e g i o n a l s t u d i e s o v e r t h e p a s t f i f t e e n y e a r s , an d r e s u l t e d i n s e l e c t i o n o f th e WIPP s tu d y a r e a f o r g e o l o g i c a l c h a r a c t e r i z a t i o n . T he w ork o f g e o l o g i c a l c h a r a c t e r i z a t i o n s h o u ld be c o n s id e r e d t o h av e b eg u n w i th t h e d r i l l i n g o f ERDA 9 , a t t h e c e n t e r o f t h e WIPP s i t e , an d t h e i n i t i a t i o n o f s e i s m ic r e f l e c t i o n s u r v e y s a t t h e s i t e i n e a r l y 1 9 7 6 . T h a t g e o l o g i c a l c h a r a c t e r i z a t i o n , w h ic h i s p r i m a r i l y o r i e n t e d t o p r o v id e s p e c i f i c d a t a c o n c e r n in g t h e p r e s e n t g e o lo g y o f th e s i t e , w as v i r t u a l l y c o m p le te i n 1978 when t h e G e o lo g i c a l C h a r a c t e r i z a t i o n R e p o r t [1 ] w as s u b m i t t e d t o t h e D e p a r tm e n t o f E n e rg y ; much b a s i c i n f o r m a t i o n h a s b e e n g a t h e r e d i n d i c a t i n g no m a jo r t e c h n i c a l p ro b le m s w i th t h e s i t e a s i t i s now u n d e r s to o d . S t u d i e s o f l o n g - t e r m p r o c e s s e s w h ic h m ig h t a f f e c t a r e p o s i t o r y o r ' h av e an e f f e c t on s a f e t y a n a l y s e s a r e now th e m a jo r g e o t e c h n i c a l a c t i v i t y f o r t h e WIPP p r o j e c t . T h e se s t u d i e s c o n c e r n t h e ag e o f s i g n i f i c a n t f e a t u r e s an d t h e r a t e s an d p r o c e s s e s w h ich p r o d u c e t h o s e f e a t u r e s . The i n f o r m a t i o n so g a in e d w i l l be u s e f u l i n i n c r e a s i n g th e c o n f id e n c e i n e v a l u a t i o n o f t h e s a f e t y o f a r e p o s i t o r y when a d e c i s i o n i s n e c e s s a r y r e g a r d i n g c o n v e r s io n o f th e WIPP t o a r e p o s i t o r y .
5 . EXPLORATION TECHNIQUES
Much o f t h e g e o l o g i c a l c h a r a c t e r i z a t i o n o f t h e WIPP s tu d y a r e a h a s em p lo y e d e x p l o r a t i o n g e o p h y s ic s an d b o r e h o l e s . A bou t 145 l i n e m i l e s o f new s e i s m i c r e f l e c t i o n d a t a an d 9 000 r e s i s t i v i t y m e a s u re m e n ts w ere c o l l e c t e d and 55 d r i l l h o l e s h av e b e e n c o m p le te d t o s u p p o r t WIPP g e o l o g i c a l c h a r a c t e r i z a t i o n t o d a t e (M ay, 1 9 7 9 ) .
A b o u t 1 500 l i n e m i l e s o f s e i s m ic r e f l e c t i o n d a t a a v a i l a b l e f ro m p e t r o l e u m c o m p a n ie s an d 26 l i n e m i l e s i n i t i a l l y o b t a i n e d s t r i c t l y f o r t h e s tu d y a r e a w e re c o l l e c t e d u s in g s t a n d a r d t e c h n i q u e s f o r t h e p e t r o l e u m i n d u s t r y . The d a t a a r e e x c e l l e n t f o r i n t e r p r e t i n g d e e p e r s t r u c t u r e , b u t a r e l e s s u s e f u l f o r sh o w in g r e f l e c t o r s i n t h e u p p e r 3 000 f e e t . I n 1977 an d 1 9 7 8 , a b o u t 1 2 0 l i n e m i l e s o f new d a t a w e re c o l l e c t e d u s in g s h o r t e r s p a c in g s f o r g e o p h o n e s , h ig h e r f r e q u e n c i e s f ro m v i b r o s e i s u n i t s , an d h ig h e r r a t e s o f d a t a s a m p l in g . T h e se d a t a show much
278 HILL
R 31 E
TD*Total Depth
LEGENDS E R D A Potash Drill Holes
(Р/ — P¿/ )ТА = Temporarily Abandoned© „ _ , л ////////// State Land
Deep Producing Gasj ■ .•> .i - —■ <—■ —• Natural Gas Pipeline
Abandoned Well. . . . . ------------ Land Withdrawal Boundary
Deep and Abandoned 'Ф Potash Drill Holes
О Geological Holes
* Hydrological Holes a)1 acre-4.047 x103 m2
ZONE AREA
I, , a)БВ acres
П 1 860 "
Ш 6 230 "
I E 10 812 "
TOTAL 18 980 acres
FIG.3. Map o f WIPP site.
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im p ro v e d r e f l e c t i o n s f ro m , and b e t t e r r e s o l u t i o n i n , t h e s h a l lo w s e c t i o n o f i n t e r e s t . R e s i s t i v i t y h a s a l s o b ee n e x t e n s i v e l y u s e d a s a c h a r a c t e r i z a t i o n t o o l . F i e l d t e s t s i n d i c a t e d t h a t r e s i s t i v i t y c o u ld d e t e c t c e r t a i n t y p e s o f s o l u t i o n f e a t u r e s ; m ore th a n 9 000 m e a s u re m e n ts h a v e b e e n t a k e n i n t h e s tu d y a r e a t o s e a r c h f o r s u c h f e a t u r e s . A d d i t i o n a l m e a s u re m e n ts o f r e s i s t i v i t y u s in g e x p a n d e r a r r a y s h a v e b e e n m ade t o s tu d y r e s i s t i v i t y c h a n g e s w i th d e p th an d t o h e lp i n t e r p r e t t h e d e t a i l e d m e a s u r e m e n ts . One r e s i s t i v i t y an o m aly was d r i l l e d t o d e t e r m in e th e c a u s e o f th e a n o m a ly . T h is a n o m aly d i d n o t r e s u l t fro m s a l t d i s s o l u t i o n phenom ena an d i s o f no c o n s e q u e n c e t o s i t e a c c e p t a b i l i t y . F u r t h e r d e t a i l e d g e o p h y s i c a l i n t e r p r e t a t i o n s o f t h e s i t e , u s in g t e c h n iq u e s f o r b e t t e r r e s o l u t i o n o f s h a l lo w h o r i z o n s , a r e now u n d e r w ay .
T h i r t e e n g e o l o g i c e x p l o r a t o r y h o le s h a v e b e e n d r i l l e d t o d a t e i n s u p p o r t o f t h i s p ro g ra m ; t h r e e h o le s w ere d r i l l e d a t t h e o l d s tu d y a r e a , tw o a r e l o c a t e d o f f t h e WIPP s i t e , and e i g h t w ere d r i l l e d on th e WIPP s i t e , F ig u r e 3 . T h e se b o r e h o le s w ere e x t e n s i v e l y c o r e d , lo g g e d , an d d r i l l - s t e r n t e s t e d f o r t h e p r o d u c t i o n o f f l u i d s . The c o r e s fo rm th e b a s i s f o r s e v e r a l c o n t i n u i n g l a b o r a t o r y s t u d i e s t h a t a r e im p o r t a n t t o an u n d e r s t a n d in g o f t h e p h y s i c a l an d c h e m ic a l pheracm ena a s s o c i a t e d w i th t h e WIPP and c o n t r i b u t e t o g e n e r a l k n o w led g e a b o u t th e f o r m a t io n o f e v a p o r i t e s . Two o f th e e x p l o r a t o r y b o r e h o le s have b e e n d r i l l e d w e l l o u t s i d e t h e im n e d ia te s i t e t o o b t a i n d i s s o l u t i o n an d p a l e o c l i m a t e d a t a . S ix h o l e s h av e b e e n d r i l l e d i n N a sh D raw , a d i s s o l u t i o n an d e r o s i o n f e a t u r e w e s t o f th e s i t e . T w e n ty -o n e h o le s w ere d r i l l e d i n c o n fo rm a n c e w i th i n d u s t r y s t a n d a r d s t o o b t a i n c o r e f ro m t h e p o t a s h z o n e s t o s u p p le m e n t m ore th a n 30 e x i s t i n g i n d u s t r y h o l e s i n e v a l u a t i o n o f p o t a s h r e s o u r c e s w i t h i n t h e WIPP s tu d y a r e a [6 , 7 ] . F o u r te e n h y d r o l o g i e h o le s h av e b ee n d r i l l e d an d f o u r p o t a s h h o le s c o n v e r t e d t o h y d r o lo g ie m o n i t o r i n g t o p r o v id e a t o t a l o f e i g h t e e n h o le s now d e d i c a t e d t o h y d r o lo g ie s t u d i e s . H y d ro lo g ie t e s t s o f t h e B e l l C an y o n F o rm a tio n u n d e r ly i n g t h e e v a p o r i t e s h a v e a l s o b e e n c o n d u c te d i n t h r e e o f th e e x p l o r a t o r y b o r e h o l e s , tw o n o r t h e a s t an d o n e s o u th o f t h e s i t e . E x c e p t f o r ERDA 9 , n o n e o f t h e b o r e h o le s w i t h i n Z ones I , I I , o r I I I p e n e t r a t e a s d e e p a s t h e r e p o s i t o r y h o r i z o n s , F ig u r e 3 .
6 . GEOLOGICAL SETTING
S t u d i e s o f t h e r e g i o n a l g e o lo g y h a v e p r o v id e d a b r o a d a s s e s s n e n t o f t h e s u r f a c e an d s u b s u r f a c e e n v iro n m e n t o f t h e a r e a w i t h i n a r a d i u s o f a b o u t 200 m i l e s o f t h e p r o p o s e d WIPP s i t e . T h e s e s t u d i e s c o n t a i n a s y n t h e s i s o f t h e a v a i l a b l e d a t a an d r e s u l t s o f r e c e n t f i e l d p ro g ra m s p e r t a i n i n g t o th e p h y s io g r a p h y an d g e c m o rp h o lo g y ,' s t r a t i g r a p h y a n d l i t h o l o g y ,
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structure, tectonic development and geologic history of this region. Such information is necessary to understand the geological processes that need to be understood for assessment of long-term safety of a repository in the Delaware Basin of southeastern New Mexico.
The r e g i o n a l g e o lo g y show s t h a t t h e n o r t h e r n D e la w a re B a s in h a s b e e n a p a r t o f a l a r g e s t r u c t u r e r e a c t i n g s lo w ly t o t e c t o n i c an d c l i m a t i c p r o c e s s e s . A b o u t 300 m i l l i o n y e a r s o f P a l e o z o i c g e o l o g i c h i s t o r y i n d i c a t e a d o w nw arp ing b a s i n on a l a r g e s c a l e . The l a s t 200 m i l l i o n y e a r s a r e c h a r a c t e r i z e d by s le w u p l i f t r e l a t i v e t o s u r r o u n d i n g s r e s u l t i n g i n sem e e r o s i o n an d d i s s o l u t i o n o f r o c k s i n th e D e la w a re B a s in . D ra m a t ic g e o l o g i c e v e n t s s u c h a s f a u l t s an d v o l c a n i c a c t i v i t y h a v e n o t o c c u r r e d i n th e n o r t h e r n D e la w a re B a s in w h ere t h e WIPP s i t e i s l o c a t e d . The n e a r e s t e v e n t s o f t h i s ty p e a r e o c c u r r i n g w e s t o f t h e G u a d a lu p e M o u n ta in s a b o u t 70 m i l e s s o u th w e s t o f t h e WIPP s i t e . T he r e g i o n a l g e o lo g y d o e s n o t i n d i c a t e t h a t an y d r a m a t ic c h a n g e s i n g e o l o g i c p r o c e s s e s o r r a t e s h av e r e c e n t l y o c c u r r e d a t t h e WIPP s i t e .
Much i n v e s t i g a t i v e e f f o r t h a s b e e n e x p e n d e d t o d e f i n e s u b s u r f a c e g e o l o g i c c o n d i t i o n s a t t h e WIPP s i t e . T h e se s t u d i e s n o t o n ly p r o v id e d e t a i l e d i n f o r m a t i o n r e g a r d i n g m in in g c o n d i t i o n s a t t h e r e p o s i t o r y l e v e l s , , b u t a l s o f u r n i s h a b a s i s f o r an a s s e s s m e n t o f th e l e v e l o f p r o t e c t i o n o r s a f e g u a r d a g a i n s t p o s s i b l e m odes o f c o n ta in m e n t f a i l u r e a t t h e s i t e , i,n t h e c o n t e x t o f th e l o n g - t e r m i s o l a t i o n r e q u i r e m e n ts o f r a d i o a c t i v e w a s te . I n f o r m a t io n fro m g e o l o g i c i n v e s t i g a t i o n s c o n d u c te d by th e U .S . G e o lo g i c a l S u rv e y a s w e l l a s d e t a i l s o f s a l t d e f o r m a t io n i n v e s t i g a t e d by o th e r c o n s u l t a n t s , h a v e d o n e much t o d e f i n e th e g e n e r a l g e o l o g i c c o n d i t i o n s i n t h e v i c i n i t y o f t h e WIPP. From d a t a o b t a i n e d by g e o p h y s i c a l t e c h n iq u e s and d r i l l i n g , a s e r i e s o f s t r u c t u r e c o n to u r an d i s o p a c h m aps h av e b e e n c o n s t r u c t e d . A c o n t i n u o u s l y c o r e d s t r a t i g r a p h i e t e s t h o l e , ERDA-9, h a s b e e n d r i l l e d t o a d e p th o f 2 864 f e e t b e lo w g ro u n d s u r f a c e n e a r t h e c e n t e r o f t h e WIPP s i t e an d p r o v id e s m uch d e t a i l on g e o l o g i c p r o p e r t i e s a t an d ab o v e t h e WIPP r e p o s i t o r y l e v e l s .
The s u r f a c e o f t h e WIPP s i t e i s a p l a i n s l o p i n g g e n t l y s o u th w e s t a t a b o u t 50 f e e t p e r m i l e ; e l e v a t i o n s r a n g e a t t h e s i t e f r o n a b o u t 3 300 t o 3 600 f e e t ab o v e s e a l e v e l . T h e re a r e no p e rm a n e n t d r a in a g e c o u r s e s i n t h e s i t e a r e a . To t h e w e s t , N a sh Draw i s a b ro a d s w a le o f a b o u t 150 f e e t o f r e l i e f l e a d i n g s o u th w e s t to w a rd t h e P e c o s R i v e r . A s l o p e m a rk in g t h e e a s t e d g e o f N ash Draw i n t h e s i t e a r e a , known a s L i v i n g s t o n R id g e , i s l o c a t e d a b o u t 4 m i l e s w e s t o f t h e c e n t e r o f t h e s i t e . N ash D raw , now p a r t l y f i l l e d w i th P l e i s t o c e n e s e d im e n t s , h a s e v o lv e d
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b o th by s u r f a c e e r o s i o n an d by s u b s u r f a c e d i s s o l u t i o n o f s a l t , p r e s u m a b ly d u r in g w e t t e r i n t e r v a l s i n t h e g e o l o g i c p a s t .
The p r o p o s e d WIPP u n d e rg ro u n d s t o r a g e f a c i l i t i e s a r e t o b e p l a c e d n e a r th e m id d le o f a 3 6 0 0 - f o o t - t h i c k s e q u e n c e o f r e l a t i v e l y p u r e e v a p o r i t e s t r a t a c o n t a i n i n g p r i m a r i l y ro c k s a l t an d a n h y d r i t e , l y i n g b e tw e e n d e p t h s o f a b o u t 500 a n d 4 100 f e e t b e n e a th g ro u n d s u r f a c e , F ig u r e 2 . The f o r m a t io n r i c h e s t i n r o c k s a l t , t h e S a la d o F o r m a t io n , i s n e a r l y 2 000 f e e t t h i c k an d c o n t a i n s t h e r e l a t i v e l y p u r e s a l t l a y e r s i n w h ic h t h e tw o p r o p o s e d u n d e rg ro u n d s t o r a g e l e v e l s a r e t o be c o n s t r u c t e d , a t a d e p th n e a r 2 120 f e e t f o r t h e u p p e r l e v e l an d n e a r 2 670 f e e t f o r th e lo w e r . The s t o r a g e h o r iz o n s a r e w e l l i s o l a t e d fro m th e h y d r o lo g ie e n v iro n m e n t by a d j a c e n t e v a p o r i t e s t r a t a . A t h i c k n e s s o f a t l e a s t 1 300 f e e t o f u n d i s t u r b e d e v a p o r i t e r o c k , p r i m a r i l y r o c k s a l t , o v e r l i e s t h e u p p e r s t o r a g e h o r iz o n and a b o u t an e q u i v a l e n t t h i c k n e s s o f a n h y d r i t e an d r o c k s a l t i n t e r v e n e s b e tw e e n t h e lo w e r s t o r a g e h o r iz o n an d t h e n e x t a d j a c e n t u n d e r ly i n g n o n - e v a p o r i t e f o r m a t i o n . The s a l t d e p o s i t s o f t h e C a s t i l e an d S a la d o F o rm a t io n s a t t h e WIPP s i t e w ere fo rm e d a b o u t 225 m i l l i o n y e a r s ag o an d h av e r e m a in e d i s o l a t e d f ro m d i s s o l u t i o n s i n c e a b o u t t h a t t im e .
Thé 2 0 0 0 - f o o t t h i c k n e s s o f t h e s a l t - r i c h S a la d o F o rm a tio n i s d i v i d e d i n t o th re e -m e m b e rs by t h e r e c o g n i t i o n o f a m id d le member r e f e r r e d t o a s t h e M cN utt p o t a s h z o n e , w h ic h i s th e i n t e r v a l w i t h i n t h e S a la d o t h a t c o n t a i n s t h e p o t e n t i a l r e s e r v e s o f p o t a s h , o r p o ta s s iu m m i n e r a l s c o m m e r c ia l ly m in e d i n th e C a r l s b a d d i s t r i c t w e s t o f t h e s i t e . A g e o c h r o n o lo g ic s tu d y o f r u b id iu m an d s t r o n t i u m i s o t o p i c c o n t e n t s o f p o t a s h m i n e r a l s g a v e a d a t e o f 206 m i l l i o n y e a r s i n d i c a t i n g t h e s y s te m h a s r e m a in e d f r e e fro m s o l u t i o n i n g a n d r e c r y s t a l l i z a t i o n s i n c e t h a t t im e . P o ta s s iu m /a r g o n s t u d i e s h av e c o n f i r m e d t h i s d a t e . The lo w e s t member o f th e S a l a d o , b e n e a t h t h e M cN utt p o t a s h m em ber, i s t h e m em ber t h a t c o n t a i n s t h e n e a r l y p u r e h a l i t e c h o s e n f o r th e p r o p o s e d f a c i l i t y . The C a s t i l e F o rm a t io n b e n e a th th e S a la d o c o n t a i n s h i g h l y p u r e b e d s o f h a l i t e b u t , u n l i k e t h e S a l a d o , a l s o c o n t a i n s t h r e e m a s s iv e a n h y d r i t e b e d s .
B a sed on in f o r m a t i o n g a t h e r e d t o d a t e , a c t i v e t e c t o n i c f a u l t i n g an d w a rp in g o f r o c k s i n t h e s i t e v i c i n i t y seem s t o h a v e p r e d a t e d P e rm ia n e v a p o r i t e d e p o s i t i o n ; c e r t a i n m in o r f a u l t i n g w i t h i n t h e t h i c k P e rm ia n s e c t i o n a p p e a r s t o h a v e o c c u r r e d c o n te m p o ra n e o u s ly w i t h s e d i m e n t a t i o n an d may be a s c r i b e d t o c o m p a c t io n . D e fo rm a tio n r e l a t e d t o s a l t f lo w a g e h a s o c c u r r e d p r i m a r i l y i n t h e C a s t i l e F o r m a t io n b e n e a th t h e S a l a d o , an d h a s l o c a l l y m o d i f i e d t h e r e g i o n a l e a s t e r l y g r a d i e n t t o 80 t o 1 0 0 f e e t p e r m i l e a t th e l e v e l o f th e s t o r a g e h o r iz o n s n e a r t h e b a s e o f t h e S a la d o . A re a s i n t h e v i c i n i t y o f t h e s i t e
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i n w h ich a r t e s i a n b r i n e r e s e r v o i r s h av e b ee n e n c o u n te r e d a r e a s s o c i a t e d w i th t h i c k e n e d s a l t s e c t i o n s an d s a l t - f l o w a n t i c l i n e s i n th e C a s t i l e , b u t no s u c h m a jo r s t r u c t u r a l f e a t u r e s a r e r e c o g n i z a b l e w i t h i n t h e l i m i t s o f t h e WIPP s t o r a g e f a c i l i t y . The s i t e a p p e a r s t o b e i n a s l i g h t s t r u c t u r a l s a d d l e , a c o n d i t i o n c o n s id e r e d t o b e f a v o r a b l e f o r s i t e s e l e c t i o n . D i s s o l u t i o n o f b e d d e d s a l t a t th e s i t e h a s b ee n r e s t r i c t e d t o h o r iz o n s w i t h i n t h e R u s t l e r F o rm a tio n a b o v e t h e S a la d o ; t h e r e i s no e v id e n c e t h a t th e r e s u l t i n g s e t t l e m e n t p r o d u c e d an y s i g n i f i c a n t s t r u c t u r a l i r r e g u l a r i t i e s o r c o l l a p s e f e a t u r e s i n th e o v e r l y i n g s t r a t a w i t h i n th e a r e a o f t h e Los M edaños s i t e . I n v e s t i g a t i o n s a r e c o n t i n u i n g t o f u r t h e r d e f i n e th e e x t e n t t o w h ich s a l t d e f o r m a t io n i n t h e C a s t i l e may h a v e a f f e c t e d t h e s t r u c t u r a l c o n f i g u r a t i o n w i t h i n t h e lo w e r p a r t o f t h e S a la d o w h ere e x c a v a t io n o f th e u n d e rg ro u n d w o rk in g s f o r WIPP i s p r e s e n t l y p la n n e d . T h e s e i n v e s t i g a t i o n s w i l l p e r m i t a m ore d e t a i l e d a s s e s s m e n t o f th e op tim um l a y o u t , d e s ig n and c o n s t r u c t i o n m e th o d o f t h e s t o r a g e f a c i l i t y .
7 . HYDROLOGIC SETTING
H y d r o lo g ie s t u d i e s o f th e p ro p o s e d s i t e an d a d j a c e n t a r e a a r e d i r e c t e d to w a rd a m ore q u a n t i t a t i v e e v a l u a t i o n o f t h e s a l t d i s s o l u t i o n p r o c e s s , t h e h y d r o g e o lo g ic p a r a m e te r s a f f e c t i n g g r o u n d w a te r m o v em en t, an d t h e m a jo r e le m e n ts o f s u r f a c e and g ro u n d w a te r q u a l i t y a s r e l a t e d t o w a te r r e s o u r c e u s e an d l o c a l e c o lo g y . The c o l l e c t i o n o f h y d r o lo g ie d a t a i s p r o j e c t e d t o c o n t in u e f o r s e v e r a l y e a r s t o p r o v id e s i t e - s p e c i f i c i n f o r m a t i o n f o r a d e t a i l e d s a f e t y a n a l y s i s o f t h e WIPP.
The o n ly m a jo r s t r e a m n e a r t h e s i t e i s th e P e c o s R iv e r w h ic h f lo w s s o u t h e a s t e r l y th r o u g h C a r l s b a d , F i g u r e 1 . A t i t s c l o s e s t p o i n t , t h e r i v e r i s a p p r o x im a te ly 14 m i l e s s o u th w e s t o f t h e WIPP s i t e . The maximum r e c o r d e d f l o o d on t h e P e c o s R iv e r o c c u r r e d o n A u g u s t 2 3 , 1 9 6 6 , w i t h a d i s c h a r g e o f 120 000 c u b i c f e e t p e r s e c o n d an d a maximum w a te r s u r f a c e e l e v a t i o n o v e r 300 f e e t b e lc w th e minimum s u r f a c e e l e v a t i o n o f th e s i t e .
C l i m a t o l o g i c a l r e c o r d s show t h a t m ean a n n u a l p r e c i p i t a t i o n a t th e s i t e i s a p p r o x im a te ly 12 in c h e s p e r y e a r . S u r f a c e d r a in a g e p a t t e r n s a t t h e s i t e a r e u n d e v e lo p e d . W a te r l e a v e s t h e s u r f a c e q u i c k l y du e t o h ig h e v a p o r a t i o n an d i n f i l t r a t i o n r a t e s . S a n d y , g r a v e l l y s o i l s c o v e r t h e r e g i o n . The n e a r e s t known g r o u n d w a te r i s m ore th a n 50 f e e t b e lo w t h e l a n d s u r f a c e . A lth o u g h t h e s e w a te r s a r e p o t a b l e , th e y a r e n o t fo u n d f r e q u e n t l y . D ee p er g ro u n d w a te r s w i t h i n t h e D e la w a re B a s in a r e p r e d o m in a n t ly o f p o o r q u a l i t y , w i th t o t a l d i s s o l v e d s o l i d s c o n c e n t r a t i o n s t y p i c a l l y i n e x c e s s o f 3 000 ppm.
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I n t h e WIPP s tu d y a r e a , e i g h t e e n h y d r o lo g ie h o le s a r e b e in g u s e d t o t e s t an d m o n i to r t h r e e f l u i d - b e a r i n g z o n e s ab o v e th e S a la d o s a l t , an d t h r e e d e e p h o le s h a v e b ee n u se d t o t e s t f l u i d sb e lo w th e e v a p o r i t e s . Above t h e s a l t , t h e u n i t s i n c l u d e tw od o lo m i te b e d s i n t h e R u s t l e r F o rm a tio n an d t h e R u s t l e r - S a l a d o c o n t a c t z o n e , s e e F i g u r e 2 . T r a n s m i s s i v i t i e s i n t h e m o s t p r o d u c t i v e d o lo m i te b ed (28 f e e t t h i c k ) v a r y f ro m 140 f t ^ / d a y a t t h e w e s t e r n b o u n d a ry o f t h e s tu d y a r e a , h o le P - 1 4 , t o l e s s t h a n 10- 4 f t 2/ d a y a t h o le P -1 8 t o t h e e a s t , F ig u r e 3 . The c o n t a c t zo n e h a s m uch l e s s p o t e n t i a l t o t r a n s m i t f l u i d s , w i th ay i e l d o f 9 g a l l o n s 1 i n a 2 0 -h o u r p e r i o d b e in g th e maximumr e c o r d e d . P o t e n t i a m e t r i c h e a d s o f a num ber o f t h e s e w e l l s h a v e n o t s t a b i l i z e d i n 22 m o n th s o f m e a s u re m e n t. P r o d u c t io n fro m w e l l s b e lo w t h e s a l t h a s b e e n l e s s t h a n 30 g a l l o n s p e r m in u te w i t h s t a t i c p o t e n t i o m e t r i c h e a d s l e s s th a n t h o s e a t t a i n e d by t h e s h a l lo w e r f l u i d s a b o v e t h e s a l t .
A num ber o f p ro g ra m s a r e b e in g c o n d u c te d t o e x t e n d t h e d a t a b a s e on t h e h y d r o l o g i e s y s te m i n an d a ro u n d t h e WIPP s i t e an d t o r e f i n e c u r r e n t u n d e r s t a n d in g o f s a l t d i s s o l u t i o n p r o c e s s e s . H o le s a r e b e in g d r i l l e d f u r t h e r f ro m t h e s i t e t o o b t a i n r e g i o n a l h y d r o lo g ie i n f o r m a t i o n . F i e l d i n v e s t i g a t i o n s o f d i s s o l u t i o n a r e f o c u s e d l o c a l l y o n f e a t u r e s n e a r t h e C a p i t a n R e e f c a l l e d " b r e c c i a p ip e s " an d r e g i o n a l l y on th e t o p o f s a l t . The fo rm e r a r e com posed o f d i s p l a c e d , r e c e m e n te d , s e d im e n ta r y r o c k ; t h e p ro g ra m w i l l d e t e r m in e i f t h e s e f e a t u r e s p o s e a t h r e a t t o th e p r o p o s e d W IPP. R e g io n a l d i s s o l u t i o n r a t e s h av e b e e n e s t i m a t e d a t a b o u t 300 f e e t p e r m i l l i o n y e a r s v e r t i c a l l y a n d 6 t o 8 m i l e s p e r m i l l i o n y e a r s h o r i z o n t a l l y . A t t h e s e e s t i m a t e d r a t e s , t h e f a c i l i t y w o u ld r e m a in s e c u r e f o r o v e r 4 m i l l i o n y e a r s .
8. ENERGY AND MINERAL RESOURCES
Geologic studies related to site characterization have included investigation of mineral resources so that an evaluation could be made of the impact of denying access to these resources. Of the mineral resources expected to occurbeneath the site, potash salts and natural gas may be ofeconomic significance. Other minerals and elements are present but are not likely ever to be of economic significance.
P o ta s s iu m s a l t s o c c u r i n a v a r i e t y o f m in e r a l t y p e s , b u t o n ly s y l v i t e (KC1) a n d l a n g b e i n i t e [ K2Mg2 (SO4 ) 3 ] a r e m in e d i n t h e C a r l s b a d P o ta s h M in in g D i s t r i c t , w h ic h i s t h e l a r g e s t s o u r c e o f p o t a s h i n t h e U n i te d S t a t e s , a c c o u n t in g f o ra p p r o x im a te ly 40% o f US u s a g e an d o v e r 80% o f d o m e s t ic
1 1 gallon (US) = 3.785 X 10~3 ra3.
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p r o d u c t i o n . The US B u re a u o f M in es h a s ju d g e d t h a t a l a n g b e i n i t e d e p o s i t l o c a t e d i n t h e n o r t h e a s t q u a d r a n t o f t h e WIPP s i t e c o u ld be p r o f i t a b l y m in e d u s in g t o d a y 's te c h n o lo g y a t t h e c u r r e n t m a rk e t p r i c e f o r t h e r e f i n e d p r o d u c t . The d e p o s i t e x t e n d s b ey o n d t h e b o u n d s o f th e WIPP s i t e , w i t h a b o u t 1 .2 m i l l i o n to n s o f e s t i m a t e d l a n g b e i n i t e p r o d u c t l y i n g i n s i d e t h e WIPP w i th d r a w a l a r e a w h e re m in in g w o u ld be p r o h i b i t e d u n d e r c u r r e n t p r e s c r i p t i o n s , Z o n es I , I I an d I I I show n i n F ig u r e 3 . S e v e r a l d e p o s i t s o f s y l v i t e a r e p r e s e n t , b u t n one a r e c u r r e n t l y e c o n o m ic .
L a n g b e i n i t e i s th e m o s t s i g n i f i c a n t m in e r a l r e s o u r c e u n d e r t h e WIPP s i t e . I t i s a s p e c i a l i z e d a g r i c u l t u r a l f e r t i l i z e r t h a t f i n d s i t s u se o n c r o p s t h a t n e e d p o ta s s iu m b u t c a n n o t t o l e r a t e a d d i t i o n a l c h l o r i n e . S o u t h e a s t New M ex ico i s t h e p r im e e c o n o m ic s o u r c e f o r t h i s p a r t i c u l a r m in e r a l i n t h e f r e e w o r ld w i th an e s t i m a t e d r e s e r v e o f o v e r 40 m i l l i o n t o n s .2 L a n g b e i n i t e e q u i v a l e n t i s a l s o p r o d u c e d s y n t h e t i c a l l y from p o ta s s iu m an d m agnesium s u l f a t e s f r a n b r i n e l a k e s . The e s t i m a t e d am ount o f s y n t h e t i c l a n g b e i n i t e i n th e US i s o v e r 300m i l l i o n t o n s , m o s t ly f rc m t h e G r e a t S a l t L a k e , U ta h .
N a t u r a l g a s , a c c o m p a n ie d w i th som e d i s t i l l a t e , a n d o i l w i tha s s o c i a t e d g a s a r e b e in g p ro d u c e d f rc m v a r i o u s b e d s i n th eD e la w a re B a s in . One p a r t i c u l a r f o r m a t i o n , t h e M orrow o f P e n n s y lv a n ia n a g e , i s o f t e n a p r o d u c e r i n t h i s r e g i o n , an d t h e e x p l o r a t i o n r i s k ( " w i l d c a t t i n g " ) i s j u s t i f i a b l e i n much o f th e w e s t e r n h a l f o f t h e s i t e . A bou t 27 b i l l i o n 3c u b i c f e e t o f n a t u r a l g a s a c c o m p a n ie d by a b o u t 0 .5 m i l l i o n b a r r e l s o f d i s t i l l a t e a r e e s t i m a t e d to be e c o n o m ic a l ly r e c o v e r a b l e f ro m b e n e a th t h e WIPP s i t e . N a tu r a l g a s p r o d u c t i o n i s g e n e r a l l y f rc m d e p th s o f 10 000 t o 14 000 f e e t . T h is am o u n ts t o a b o u t0.01% o f th e c u r r e n t l y e s t i m a t e d US r e s e r v e .
9 . CONTINUING STUDIES
A lth o u g h much d e t a i l e d i n f o r m a t i o n h a s b ee n r e p o r t e d a s a r e s u l t o f s i t e s e l e c t i o n and c h a r a c t e r i z a t i o n , t h e r e re m a in a num ber o f p ro g ra m s w h ich h av e n o t oome t o c o m p le t io n , o r i n sam e c a s e s , h a v e n o t b e g u n . T h e se p e n d in g p ro g ra m s a r e a im ed a t r e f i n i n g a n d s u p p le m e n t in g in f o r m a t i o n g a t h e r e d t o i n c r e a s e t h e c o n f id e n c e to be p l a c e d i n f a c t o r s r e l a t i n g t o s i t e s e l e c t i o n . F u r th e r m o r e , som e k in d s o f i n f o r m a t i o n r e m a in t o be g a t h e r e d t o s u p p o r t l a b o r a t o r y an d i n s i t u e x p e r im e n t s a n d lo n g - t e r m s a f e t y a s s e s s m e n t .
2 Short tons = 9.072 X 102 kg.3 US billion = 10®
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S an e o f th e p ro g ra m s w h ich p r o v id e t e c h n i c a l s u p p o r t f o r t h e d e v e lo p m e n t an d s a f e t y a s s u r a n c e o f t h e WIPP a r e d e s c r i b e d b e lo w .
1 . b r i n e m i g r a t i o n - c h a r a c t e r i z a t i o n o f th e b e h a v io r o f n a t u r a l l y o c c u r r i n g f l u i d i n c l u s i o n s i n e v a p o r i t e r o c k s u n d e r th e i n f l u e n c e o f th e rm a l g r a d i e n t s .
2 . c o r r o s i o n - d e t e r m i n a t i o n o f w a s te an d w a s te c o n t a i n e r i n t e r a c t i o n s w i th r o c k s a l t u n d e r v a r i o u s p r e s s u r e , t e m p e r a t u r e an d w a te r c o n t e n t c o n d i t i o n s .
3 . th e r m a l s t r u c t u r a l i n t e r a c t i o n - d e v e lo p m e n t and a p p l i c a t i o n o f e x p e r im e n t a l d a t a , c o n s t i t u t i v e m o d e ls an d c o m p u te r c o d e s t o p r e d i c t ro c k t e m p e r a t u r e , s t r e s s and d e f o r m a t io n f i e l d s .
4 . b o r e h o le p lu g g in g - s e l e c t i o n an d t e s t i n g o f c e m e n t i t i o u s g r o u t s f o r p lu g g in g b o r e h o le s an d s h a f t s b a s e d on p e r m e a b i l i t y , bond s t r e n g t h an d lo n g - te r m s t a b i l i t y c r i t e r i a .
E ach o f t h e s e p ro g ra m s i s b e in g a d d r e s s e d u s in g d i f f e r i n g s i z e d s a m p le s , p r o g r e s s i n g f ro m l a b o r a t o r y a n a ly s e s t o f u l l s c a l e i n s i t u e x p e r i m e n t s . The c o l l e c t i o n o f e x p e r im e n t a l r e s u l t s f rc m t h e s e p ro g ra m s w i l l fo rm t h e t e c h n i c a l b a s e s f o r t h e i n s i t u e x p e r im e n t a l p o r t i o n o f th e WIPP f a c i l i t y . I n f o r m a t io n g a in e d f ro m t h e s i t e c h a r a c t e r i z a t i o n and e x p e r im e n t a l p ro g ra m s w i l l be u se d t o c o n f irm h y p o th e s e s an d v a l i d a t e m o d e ls t o p r o v id e g r e a t e r c o n f id e n c e i n g e o l o g i c i s o l a t i o n o f r a d i o a c t i v e w a s t e s .
REFERENCES
[ l ] POWERS, D. W ., e t a l . , " G e o lo g ic a l C h a r a c t e r i z a t i o nR e p o r t W a s te I s o l a t i o n P i l o t P l a n t (W IPP), S o u t h e a s t e r n New M e x ic o ," SAN D 78-1596, S a n d ia L a b o r a t o r i e s ,A lb u q u e rq u e , NM ( 1 9 7 8 ) .
[ 2 ] D e p a r tm e n t o f E n e r g y , D r a f t R e p o r t o f T a sk F o rc e f o rR ev iew o f N u c le a r W a s te M anagem ent: D O E/ER-0004/D ( 1 9 7 8 ) .
[ 3 ] WIPP C o n c e p tu a l D e s ig n R e p o r t , SAND77-0274, S a n d ia L a b o r a t o r i e s , A lb u q u e r q u e , NM ( 1 9 7 7 ) .
[ 4 ] NAS/NRC, " D is p o s a l o f R a d io a c t iv e W a s te s on L a n d ," N a t i o n a l A cadem y o f S c i e n c e s - N a t i o n a l R e s e a r c h C o u n c i l , W a s h in g to n , D .C . , P u b l . 519 ( 1 9 5 7 ) .
[5 ] D r a f t E n v ir o n m e n ta l Im p a c t S ta te m e n t W a s te I s o l a t i o n P i l o t P l a n t , D O E /E IS -0 0 2 6 -D , ( 1 9 7 9 ) .
[ 6 ] JOHN, C . B . , R . J . C h e ese m a n , J . C . L o re n z , an d M. L. M i l l g a t e , " P o ta s h O re R e s e r v e s i n t h e p ro p o s e d W a ste I s o l a t i o n P i l o t P l a n t A re a , E ddy C o u n ty , S o u t h e a s t e r n New M e x ic o ," O p e n - F i le R e p o r t 7 8 -8 2 8 , US G e o lo g ic a l S u rv e y (1 9 7 8 ) .
[ 7 ] USBM (US B u re a u o f M in e s ) ," V a lu a t io n o f P o ta s h O c c u r r e n c e s W i th in t h e W a ste I s o l a t i o n P l a n t S i t e in S o u t h e a s t e r n New M e x ic o ," p r e p a r e d f o r th e US E n e rg y R e s e a r c h an d D e v e lo p m e n t A d m i n i s t r a t i o n ( 1 9 7 7 ) .
286 HILL
DISCUSSION
V.I. SPITSYN: Will you inject liquid waste or will you store the waste in
solidified form?
L.R. HILL: Only solidified wastes will be received.
V.I. SPITSYN: What material will you use for the canisters?
L.R. HILL: The canister materials are yet to be selected. Various materials
for both the canisters and waste matrices are being considered by the Waste
Acceptance Steering Committee in the USA; a report is due within a year. Some
metallic alloys look promising.
L.J. ANDERSEN: What in situ methods have been used to determine the
vertical permeability of the water-bearing and confining (impervious) beds?
L.R. HILL: So far, vertical boreholes are the only means we have to test
permeability. Vertical permeabilities are believed to be very slight because
production from the water-bearing zones is very slight — as little as one gallon
every 4 to 400 hours, and the hydraulic heads and chemistries of waters from
adjacent units differ. This conclusion is confirmed by the pressure (slug) testing
of these units. Within the salt beds, air pressure tests in a borehole indicate very
low permeabilities — less than a microdarcy.
H. KRAUSE: You mention in your paper that the area where WIPP is to be
located contains some potash salt and hydrocarbon resources. Do you think this
could affect acceptance of the site?
L.R. HILL: The question of these resources is being debated and it may have
an adverse effect on acceptance. The value of the resources is very low from the
point of view of the United States economy; however they are very valuable to
the owner of the particular lease. When the site was chosen for characterization,
we believed the resources there to be insignificant. Now that some have been
discovered we must determine what would be the jeopardy to repository integrity
IAEA-SM-243/38 287
if the resources were exploited. This is the motivation for the programmes on
subsidence and directional drilling which I described.
Valentina BALUKOVA: It is known that dry storage of radioactive wastes
in natural salt leads to the accumulation of energy as a result of radiation. What
provision is made for safety when such energy is released rapidly and, in case of
accidental entry of water, causes generation of hydrogen?
L.R. HILL: Storage of significant amounts of energy in salt by this mechanism
requires a substantial radiation source. WIPP is planned primarily for low-level
transuranic wastes, where such a radiation field would not be present. If the high-
level waste experiments are performed and this phenomenon does occur, it will
be evaluated and considered in our long-term risk assessment programme.
IAEA-SM-243/151
SITE SELECTION, SITE INVESTIGATIONS AND DESIGN ACTIVITIES IN THE USA FOR NUCLEAR WASTE REPOSITORIES IN BEDDED AND DOME SALT FORMATIONS
M. KEHNEMUYI, S.C. MATTHEWS
Battelle Memorial Institute,
Columbus, Ohio,
United States of America
Abstract
SITE SELECTION, SITE INVESTIGATIONS AND DESIGN ACTIVITIES IN THE USA FOR NUCLEAR WASTE REPOSITORIES IN BEDDED AND DOME SALT FORMATIONS.
This paper discusses the site selection criteria by which repository sites will be qualified for placement of nuclear waste repositories in deep geologic formations, and the design of repositories in bedded and dome rock salt formations. The four major rocksalt regions that are now being considered for possible placement of repositories are described and the geologic, environmental and socioeconomic criteria by which sites will be screened and qualified are described. The screening criteria fall into two categories: those that are based on legal constraints and are therefore exclusionary, and those that depend on discretionary considerations. Both these types of criteria are discussed. The design of repositories in bedded and dome salt formations is described, and the factors to be considered in site selection which would assure that the design provides a safe containment of the high-level nuclear waste are identified.
Site Selection Criteria
In the United States, at the present, several geologic formations are being investigated for locating repositories for disposal of high-level nuclear waste and spent-fuel assemblies from commercial nuclear power plants. The formations being investigated include rocksalt, granite, shale, basalt, and tuffs. Following a recommendation by the National Academy of Sciences in September, 1957, most of the scientific investigations in the United States have been centered on rocksalt.Four regions within the contiguous 48 states have been identified as possible locations considering the geologic aspects required for placing a repository below the surface of the earth. These four regions are the following:
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290 KEHNEMUYI and MATTHEWS
(1) Salina bedded salt formations in the states of Ohio, New York, Michigan, and Pennsylvania
(2) Gulf Interior dome salt formations found in the states of Mississippi, Louisiana, and Texas
(3) Permian bedded salt formations in the states of Texas, New Mexico, and Kansas
(4) Paradox anticline formations in the states of Utah and Colorado.
Selection of specific sites within these regions first requires establishment of geologic, hydrologie, environmental, and socioeconomic criteria. Once these criteria are established, the regions are then examined for determining if suitable locations that meet the requirements of the criteria exist. The primary objective of the criteria is to assure that in selecting a site for a repository, all considerations for protection of the health and safety of the public in the present and future generations are included, and that there will be no adverse effect on the environment.
As a first screening process, the following geologic criteria . are considered for site suitability.
(1) The top of the rocksalt formation should be at a depth below the earth's surface so that no surficial process or event may reach that depth. Glacial actions and meteorites may be possible causes of breaching the cover over the rocksalt.
(2) The rocksalt formation should have a real dimension that would be sufficient for the repository and allow an adequate buffer zone.
(3) The rocksalt formation should have enough depth to contain the repository with adequate thicknessof salt above and below the repository as a buffer.
In addition to the geologic criteria, socioeconomic and environmental criteria (which are commonly termed nongeologic criteria) are established.
The nongeologic screening criteria are categorized into two basic parts. These screening criteria are based on:
(1) Legal constraints which exclude use of certain lands for other than defined purposes. These would be called exclusionary criteria and cover such items as :
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(a) Wilderness Act of 1964 establishes wilderness as a land-use category and gives authorityto the Department of the Interior under guidelines to establish federally owned land areas as wilderness areas to be preserved in the wild state.
(b) Wild and Scenic Rivers Act of 1968 provides for the implementation of a national policy to preserve certain selected rivers in the free-flowing condition and to protect their immediate environments.
(c) The Threatened or Endangered Species Act of 1969 provides for the conservation of endangered and threatened species and of ecosystems on which these species depend.
(d) The National parks and National monuments which are administered by the National Park Service within the Department of the Interior represent an extraordinary National heritage protected by Federal law.
(e) National Historic Preservation Act of 1966 and the Archeological and Historic Preservation Act of 1974 exclude designated historical sites which cannot be moved and designated archeological sites which cannot reasonably be exploited to remove the valuable articles.
(2) Other land areas, in addition to those legally dedicated to special uses should be considered for exclusion. Examples of these discretionary considerations are:
(a) Urban places
(b) Significant groundwater use
(c) Surface water, flood plains
(d) Rugged topography
(e) Naturally formed landmarks
(f) Mineral resources
(g) Indian reservations
(h) Potentially interactive land uses, such as airports
(i) Not easily accessible areas.
292 KEHNEMUYI and MATTHEWS
Each of the four regions is examined for the geologic and nongeologic criteria mentioned above and availability of sites is illustrated by maps and overlays of exclusionary and discretionary considerations.
Site Selection Considerations
In the development of a geologic repository design to isolate nuclear waste, two basic considerations are paramount: assurance of long-term isolation of radioactive nuclides from the biosphere, and safe and efficient operation of the repository throughout the emplacement of waste. The first consideration, that of long-term isolation, is assured by the selection of a geologically stable site and maintaining the stability of the host rock formation and the surrounding strata through proper repository excavation and through limiting the repository thermal loading to acceptable values. As an additional safety factor, it may also be deemed desirable to maintain waste form stability for extended periods.
Factors to be considered in site selection include tectonic stability, history of regional seismic activity, regional hydrology, and the slope of the formation and its surrounding strata. In addition, to reduce the probability of inadvertent penetration of the repository region it is helpful if the region is devoid of natural resources that are attractive to man.
Formation stability is principally maintained, as a design function, by controlling the thermomechanical stresses induced by repository excavation and the emplacement of heat-generating waste. For purposes of simplifying the conceptual repository design process, the limiting of thermomechanical stresses can be translated into temperature and thermal loading limits.These limits are influenced by: depth of repository, physical properties of host formation and surrounding strata, repository extraction ratio, age and composition of the heat-generating waste, amount and distribution of impurities in the host formation, and hydrology of the surrounding strata. The interrelationship of these considerations must be examined in any specific repository design, as they are so closely related that changing one parameter affecting thermal loading will influence the other factors that also influence thermal loading.
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Design of Repositories
Major elements of the DOE1 Waste Management Program are addressing not only site selection through geologic exploration, but also further refinement of repository thermal design criteria currently in hand. Concurrent with these activities, the repository conceptual design process has begun and has been extensively developed. At this time, essentially completed conceptual designs are available for both dome salt and bedded salt, for both spent unreprocessed fuel and high-level waste. Numerous additional design studies have been completed for other fuel cycles and for other candidate geologic media.
In simplest terms, a geologic repository is not unlike a mine, and consists of access corridors and disposal rooms excavated deep within the formation. Various supportive structures are provided on the surface over the repository, and shafts are constructed to provide access from these structures to the disposal level. Wastes, packaged in appropriate containers, arrive at the repository on vehicles appropriate for transport, such as trucks or rail cars, over public access routes. In the event that waste-processing facilities are located at the repository site, the packaged wastes arrive at the repository facilities in transfer carriers that might not need to meet requirements placed on containers and casks used for transport over public access routes. In either case, waste-filled containers are unloaded from the carriers at the repository and prepared in surface facilities for descent to the mine level. The containers are lowered through a shaft to the disposal level, and are then transported to disposal rooms and placed therein. As disposal rooms fill to their capacity, surplus mined and/or other material can be reintroduced to the mine in order to backfill void spaces in the disposal room, effectively burying the waste.
The surface facilities that support the repository represent the only visible evidence of the repository and are usually confined within a fairly small area. These facilities include waste- receiving and handling buildings, mining-support buildings for workers, ventilation and material handling, as well as whatever infrastructure may be necessary. Shafts connect some of these facilities to the underground workings, providing access for men, materials, waste receipts, and ventilation.
The subsurface repository layout is structured to simultaneously accomplish excavation, waste emplacement, and the backfilling of disposal rooms. The layout is such that separate ventilation
1 DOE = Department of Energy.
294 KEHNEMUYI and MATTHEWS
systems support the excavation activities and the waste emplacement activities. The total area of the underground mine might include as much as 2000 acres (3 sq. miles). The depth of the mine would vary from one repository to another, depending on the type of rock being used and the thickness and depth of the seams. Generally speaking, depths between approximately 800 and a few thousand feet are considered feasible for salt repositories.
The repository will be licensed by NRC, and the facilities are designed to protect the public and operating personnel from radiation and/or contamination hazards and to contain the radioactive materials during both normal operation and emergency or accident situations. In addition, facilities containing radioactive materials would be designed to maintain their integrity under flood, earthquake, and tornado conditions.
The title to the land area underlain by the excavations will be acquired by DOE. This will ensure the safety of the repository by preventing drilling, mining, or other activities that could conceivably breach the storage formation. However, selective surface rights for this area might be leased for grazing, farming, etc. Additional restrictions will be established on subterranean activities in an area extending outward for some distance around the excavation.
After selection of the site, the repository will be constructed in an initial phase to allow an evaluation of the handling and storage operations using actual waste. This initial-phase repository would require the excavation of only a small part of the ultimately available mine area and would handle relatively small amounts of each of the waste types. Surface facilities would be approximately full scale. The initial-phase repository would provide valuable information on the mechanics and stability of the rock when subjected to such a use, and would aid in the development of operating procedures and mechanisms that would be used in the full-scale repository. Although it is not obvious that retrievability of emplaced wastes would be a requirement for the repository, wastes received during this initial phase could be handled in such a way as to be easily retrieved, by keeping the storage rooms open and installing protective liners between the waste and the rock.
Once this initial phase is satisfactorily completed, expansion of the mine will begin, and over a period of years the entire repository acreage will be developed into storage rooms and wastes emplaced in these rooms. When filled to capacity with waste, the rooms are then backfilled with compatible rock material. Although no special operations are conducted to assist retrievability, the waste can, in fact, be retrieved from these
IAEA-SM-243/151 295
backfilled regions, should safety or other considerations require this, as long as access to the mine level is maintained. When the repository fills to capacity, all shafts and remaining accesses to the mine will be plugged and sealed, and the surface facilities would be decommissioned and dismantled.
DISCUSSION
C.J.G. BARRAUD: Has the type of backfill material been finally determined?
M. KEHNEMUYI: No. We are concentrating on optimization to provide a
total system of man-made barriers which should be compatible with the natural
geologic barriers. Salt will certainly be considered as backfill material for repositories
in salt formations.
C. DAVISON: In discussing site selection criteria, you indicated that you
would be applying discretionary criteria. Have you assigned weighting factors to
these criteria and, if so, will you be trying to use these weighted criteria in your
selection analyses?
M. KEHNEMUYI: The weighting factors will have to be somewhat judgemental
in that they may not be amenable to definite mathematical treatment. It is
important that the decision basis, even though it be judgemental, should not only be
acceptable to the public but also emphasize those factors which might be more
important than others.
J. HAMSTRA: You mentioned applying the multibarrier principle to a salt
dome repository. Could you be a little more specific about what types of barrier
you have in mind?
M. KEHNEMUYI: The barriers are associated with the waste form, the filler
material in the canister, the canister material, the backfill in the placement hole
and the backfill in the repository drifts. These constitute the man-made or
engineered barriers. In addition, of course, there are the natural barriers which
are the host rock and the geosphere around the host rock, involving such matters
as hydrologie flow paths, the distance to the biosphere and the nature of the
aquifers.
The specific type of engineered barriers will be determined by optimization
procedures designed to provide adequate containment in conjunction with the
natural barriers.
C.A. HEATH: May I make a general comment here? Questions have been
asked of Mr. Kehnemuyi and Mr. Hill concerning the exact application of siting
criteria and weighting factors. I do not believe that society will necessarily accept
results obtained by an exact formula for siting, including the use of weighting
factors which scientists might like to apply.
296 KEHNEMUYI and MATTHEWS
As an example, I should point out that the potash deposits around the
proposed WIPP site referred to by Dr. Hill were discovered only as a result of work
to investigate the site in detail. Perhaps similar exploration at Gorleben will also
reveal natural resources. Some political process will be required to resolve this
issue of potential conflict between competing needs of society.
G. STOTT : The programme involving the completion of field studies and
the selection of a site by 1984 is an ambitious one. The paper describes some of
the legal barriers to site selection but, apart from the mention of Indian reservations,
it does not discuss the potential barrier of public acceptability. Will public
acceptability be a problem?
M. KEHNEMUYI: Public acceptability has to be achieved before the site is
selected and repository construction started. Indeed it will be difficult to obtain
public acceptance but achieving this is a major part of our programme.
Indian reservations are mentioned among the discretionary criteria; some
people believe that if suitable formations exist in these places and acceptance by
the population of the reservation and of the State within which the reservation
is located can be obtained, then a repository could be established on such lands.
Basically, acceptance by the public is very important.
IAEA-SM-243/48
SITE INVESTIGATIONS AND CONCEPTUAL DESIGN FOR THE REPOSITORY IN THE NUCLEAR ‘ENTSORGUNGSZENTRUM’ OF THE FEDERAL REPUBLIC OF GERMANY
H. RÜTHEMEYER
Physikalisch-Technische Bundesanstalt,
Brunswick,
Federal Republic of Germany
Abstract
SITE INVESTIGATIONS AND CONCEPTUAL DESIGN FOR THE REPOSITORY IN THE
NUCLEAR ‘ENTSORGUNGSZENTRUM’ OF THE FEDERAL REPUBLIC OF GERMANY.
Site investigations are planned to enable the Phy sikalisch-Technische Bundesanstalt and
the other institutions involved to scrutinize the suitability of the Gorleben site and its salt dome
as the nuclear ‘Entsorgungszentrum’ repository, and to provide all data necessary for detailed
site-specific planning and safe building and operation of the repository. The investigations
include deep drillings and an extensive hydrogeological programme for exploring the general
structure of the salt dome and its environment. The latter programme started on 17 April 1979.
The Bundesanstalt für Geowissenschaften und Rohstoffe (Federal Geological Survey) is in
charge of the geoscientific side of the investigations. For the site-independent conceptual design
the following requirements have been laid down: ( 1) for safety reasons retrievability is not to be
considered; (2) standard mining techniques and the special experience gained at the Asse salt
mine should be utilized as much as possible; (3) two shafts should be sufficient for the mine,
for all transport and the ventilation system; (4) different forms of waste in terms of waste type
and container shall be disposed of in different storage areas; (5) ventilated sections must permit
the shutting off of each storage area from the rest of the mine; (6 ) the mining method of retreat
working should be applied in order to explore the whole disposal area prior to the beginning of
disposal, to enable an optimal fit of the mine to the structure of the salt dome, and to move
away from the heat generated by heat-producing waste; (7) the mine works shall have a lateral
safety distance to the cap rock of 200 m and a vertical safety zone of 300 m beneath the salt
level. (8 ) all disposal areas shall be on one level; (9) salt and waste shall be transported in
different drifts, mainly in a one-way system.
1. INTRODUCTION
With the addition of the 4th Amendment to the German Nuclear Law in
1976 Parliament laid down the basic principles for the closure of the fuel cycle.
With respect to long-term storage and final disposal of radioactive waste it states
that:
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298 RÔTHEMEYER
the Federal Government has to provide facilities for long-term storage and
final disposal of radioactive waste;
the Federal Government can be assisted in its responsibilities by an institution
of private character;
The Physikalisch-Technische Bundesanstalt (PTB) acts on behalf of the
Federal Government; i.e. is responsible for building and running the
facilities;
the “polluter pays” principle is to be implemented.
For all other parts of the nuclear fuel cycle, responsibility lies with the
industry.
On 22 February 1977, Dr. Ernst Albrecht, Prime Minister of the German state
of Lower Saxony, announced that site-selection activities had shown Gorleben
to be the most suitable site to initiate a Ucensing procedure for the Nuclear
“Entsorgungszentrum” (NEZ).
Another paper at this symposium [ 1 ] elaborates on the licensing situation
and the background to NEZ’s waste disposal facilities. Putting the views of the
people and institutions concerned into a nutshell one can say that no serious
objections exist to continuing the present activities and to starting the salt dome
investigations by deep drillings.
2. SITE INVESTIGATIONS
Generally speaking, site investigations shall
enable PTB and the other institutions involved to scrutinize the suitability
of the Gorleben site and its salt dome for NEZ’s repository;
provide all data necessary for detailed site-specific planning and safe building
and operation of the repository.
Thus the investigations include deep drillings and an extensive hydro-
geological programme for exploring the general structure of the salt dome and its
environment. The Bundesanstalt fur Geowissenschaften und Rohstoffe, Federal
Geological Survey (BGR) is in charge of the geoscientific side of the investigations.
2.1. Deep drillings for salt dome investigations
The inner structure of the Gorleben salt dome is unknown. It will be
investigated by deep drillings up to a depth of about 2000 m and, after the shafts
have been sunk, from horizontal shafts underground.
IAEA-SM-243/48 299
/ / / / / / /s a lt dome ¡300 m depth) — salt dome (1000m depth!
--------- site o f plant
• deep drilling point
areas for hydrogeologicalinvestigations
FIG.l. Site investigations.
For safety reasons the number of deep drillings will be limited to between
three and seven. They will provide information about the general structure of the salt
dome and the locations most suitable for the two shafts. Thus these drillings
will play a decisive role in the time schedule for the repository.
The drillings will be performed at the north-west and south-east fringes of
the dome. Three locations have been chosen (Fig. 1); the others can only be
fixed when the results of the first drillings are available.
The programme includes the usual methods of taking drilling cores and
performing extensive bore-hole measurements to get geological, mineralogical and
rock mechanical data on the salt dome. The information obtained from these
methods will be complemented by the application of a newly developed (Prakla-
Seismos GmbH, BGR) high-frequency method: reflection measurements from a
borehole and absorption measurements between two boreholes will identify
inhomogeneous qualities of the rock up to several hundred metres away from the
boreholes.
In view of the known size of the Gorleben dome the information obtained
from the deep drilling programme will be sufficient to confirm the suitability
of the dome for all NEZ’s light-active and medium-active waste, with negligible
heat production.
The suitability for heat-producing medium- and high-active waste can only be
judged after the shafts have been sunk and the inner structure of the dome has
been explored from underground shafts by drillings and the high-frequency method
mentioned above.
The time scheduled for the three to seven deep drillings is about two years.
The licence for the first three drillings was applied for in 1977/78. Following
the outcome of the Gorleben hearing the PTB plans to start the deep drillings late
this year.
300 RÔTHEMEYER
This programme started on 17 April 1979. It will give detailed insight into
the hydrogeological aspects of the Gorleben site and will thus provide the basis
for clarifying all important issues in connection with the groundwater, especially
the possible influence on the groundwater of building and operating the repository
and, conversely, the influence of the groundwater on the repository.
The hydrogeological aspects currently being explored will provide the
necessary data and information on the overlying strata, the characteristics of the
aquifers, and the chemical and physical parameters of the groundwater, including
rate and direction of flow.
To acquire these data the programme under way includes:
(1) about 100 drillings at an average depth of 250 m;
(2) the installation of about three wells for groundwater measurements in the
vicinity of each of the above-mentioned drillings;
(3) about 450 geoelectrical surface measurements;
(4) geophysical bore-hole measurements.
The drillings cover an area of roughly 300 km2 (Fig. 1).
The number of drillings per km2 decreases with increasing distance from the
site of the NEZ. Here one drilling will be sunk per km2, the next area is explored
by one drilling per 2 km2 añd the third one by one drilling per 4 km2. Further
areas will only be explored if further information is required for achieving the
above-mentioned goals.
Though the first seven drillings and 25 wells have been made, evaluation of
the results is still under way. Their small number would in any case not permit
general conclusions to be drawn from them.
All seven drillings were sunk to the impermeable Tertiary clay formations
at depths between 120 and 270 m. The salt level and the gypsum cap rock have
not been explored yet. No hydraulic contact between groundwater layers and the
salt level area has been discovered.
The hydrogeological programme is scheduled to finish this year. Its evaluation
will of course take considerably longer. The data and information will not only be
available to the PTB, but also to other institutions officially involved in the safety
evaluation of NEZ, especially to the experts of the licensing authorities and the
project “Safety-Studies Entsorgung” [2].
3. THE PLANNING BASIS FOR THE SITE-INDEPENDENT REPOSITORY
DESIGN
As there are hardly any site-specific data for the Gorleben site available, in
order to push the project the PTB has asked an engineering group, Konsortium
2.2. Hydrogeological programme
IAEA-SM-243/48 301
Planung Endlager (KPE), Siemag-Transplan and Deilmann-Haniel, to plan a model
mine. Their planning was finished in May this year. The report proves that a
repository for all NEZ’s waste can be built and operated using technologies
presently available and conforming to requirements laid down by the mining
authorities.
Planning was based on capacity requirements drawn from NEZ’s radio
active waste in 50 years of operation, general site-independent requirements and
geological assumptions about the salt dome model “Gorleben”.
3.1. Types and quantities of radioactive waste
The following types and quantities of solid and solidified waste based on the
annual amount of waste from NEZ’s reprocessing plant were assumed for the
site-independent repository design:
(1) Light Active Waste (LAW) in 400 ltr barrels, unshielded: 11 000 barrels
per annum;
(2) LAW producing a-radiation and
(3) Medium Active Waste (MAW) in 400 ltr barrels, in a lost concrete shielding:
12000 barrels per annum;
(4) MAW in 400 ltr barrels in retrievable shielding: 5000 barrels per annum;
(5) heat-producing MAW in 400 ltr barrels in retrievable shielding:
5000 barrels per annum;
(6) High Active Waste (HAW) in steel cylinders and with retrievable shielding:
1760 cylinders per annum.
The barrels with or without a shielding are allowed to have a dose rate of
not more than 200 mrem/h at the surface and 10 mrem/h at a distance of 1 m
from the container.
3.2. General site-independent requirements
The basic requirements for the site-independent repository design are the
following:
( 1) for safety reasons retrievability is not to be considered;
(2) standard mining techniques and the special experience gained at the Asse
salt mine should be utilized as much as possible;
(3) two shafts should be sufficient for the mine, for all transport and the
ventilation system;
302 RÔTHEMEYER
0 0 ™_______ground leve l 25 m obove sec -le v e lm m ground water level ¿um
гопе requiring frozen _ aquifer-emplacement
Г - sand. grovel, day-w ater-bearing , unstable
JBO.Qmv . -m nm collapse zone with gypsum, d o v . anhydrite,
j v w m m tp r nr h rine-hearing unstnhle
shaft bottom
rock salt, compact, stable Ioccasional \ fíooom cornallite strata above 600 m depth possible)
\ Older and lounger rock salt
\ poo.Qm storage level lim it
at
salt
dome
ax
is г " Л ............... " “
FIG.2. Hypothetical shaft profile.
] 5 ground level ; 0Qm
overlying strata
"g I FIG.3. Possible storage area.
(4) different forms of waste in terms of waste type and container shall be
disposed of in different storage areas;
(5) ventilated sections must permit the shutting off of each storage area from
the rest of the mine;
(6 ) the mining method of retreat working should be applied in order to
explore the whole disposal area prior to the beginning of the disposal,
enable an optimal fit of the mine to the structure of the salt dome, and to
move away from the heat generated by heat-producing waste.
(7) the mine works shall have a lateral safety distance to the cap rock of 20 0 m
and a vertical safety zone of 300 m beneath the salt level;
(8) all disposal areas shall be on one level;
(9) salt and waste shall be transported in different drifts, mainly in a one-way
system.
IAEA-SM-243/48 303
FIG.4. General representation of the construction in the freezing shaft.
3.3. Geological assumptions
Though the inner structure of the Gorleben salt dome is unknown, some
knowledge exists of the general geological situation of the area and especially of
the shape of the dome. The latter information has been gained by earlier seismic
measurements. This information was used to make up a hypothetical shaft profile,
consisting of (Fig.2), a covering roof at about 300 m down; a collapse zone with
gypsum, clay, anhydrite; and occasional camallite strata within the first 600 m
below ground level. The storage area between 830 and 900 m down was assumed
to consist of homogeneous rock salt (Fig.3).
4. DESIGN FEATURES
In the following sections the conceptual design of the repository is described,
based on the above-mentioned assumptions and requirements. Though this model
is in principle site-independent, its flexibility allows the adjustment of its basic
design features to specific salt domes of sufficient size.
4.1. Shaft sinking
In order to guarantee the isolation of the waste from the biosphere, licensing
requirements necessitate stringent planning and control over building especially
of the two shafts. Though the exact locations of these shafts can only be fixed
after the results of the deep drillings are available, the shafts will very likely be
located on the NEZ’s area. They will be about 500 m apart. No waste will be
disposed of within a safety pillar of 300 m around them.
304 RÔTHEMEYER
------------- southern flank of saltdome a t 1000 m depth
FIG.5. Schematic plan of shafts and mine
FIG. 6. Tumble-down technique.
The hypothetical shaft profile (Fig.2) requires the frozen aquifer-emplacement
technique to be applied. The congelation tubes will be placed in holes bored
around the shaft centre at a diameter of 16 m. The external/internal diameters of
the shafts are 11 m/7.50 m. Figure 4 shows the general constructional features
of the shaft. The lower edge of the impermeable liner will be 50 m below the
salt level.
The shaft described can be considered absolutely waterproof. It can cope
with vertical movements of the rock formation and will even withstand small
horizontal displacements.
IAEA-SM-243/48 305
О 5 10 15 20 2 5 m
FIG. 1. Remote stacking technique.
0 2 4 6 8 10 m
FIG. 8. Top-loading chamber.
4.2. Driving of the mine
The mine will be driven from the two shafts, which are sunk simultaneously.
Shaft I is used for waste transport and upcast ventilation; shaft II will serve for
downcast ventilation and transport of personnel, material and salt.
After the shafts have been sunk to the storage level, the mine works will be
driven by excavating (drilling and blasting) drifts, circumventing the envisaged
disposal areas on both sides of the shafts. Only after the whole future disposal
area has been investigated (see section 2 .1.) will the disposal chambers and bore
holes be driven. Figure 5 shows a schematic plan of the shafts and the mine. The
eastern wing will be reserved as a disposal area for heat-producing waste; in the
western wing the LAW and MAW with negligible heat production will be
disposed of.
306 RÔTHEMEYER
2600 250
E Mm
ëf$À .
°.ir>Ф
0 1 2 3 4 5 6 m
FIG. 9. Vertical borehole technique.
4.3. Disposai techniques
As has already been mentioned, different disposal techniques will be used,
depending on the type of waste. With the exception of the borehole techniques all
other techniques have been developed and tested at Asse and are well-known. The
tumble-down technique will be applied for LAW (Fig.6 ). The chamber-dimensions
are 60 X 15 X 20.5 m. One chamber can take all the LAW produced in two years.
LAW-barrels containing «-producing waste and MAW-barrels in a lost concrete
shielding are disposed of by using a remote stacking technique (Fig.7). The
chamber dimensions are 100 X 15 X 25.5 m. One and a half chambers will be
needed per annum.
Top-loading chambers will take MAW transported in retrievable shielding.
0.7 chambers are needed per annum (Fig.8 ).
Whilst the above-mentioned techniques are based on many years of
experience in the Asse mine, the borehole technique (Fig. 9) needs further R & D
work before it can be applied on an industrial scale.
The technique has to be used for heat-producing MAW and HAW. The
MAW barrels contain sludge suspension, cladding, core parts and scrap from the
vitrification process. The heat production ranges from about 100 to 300 W/barrel.
These barrels are disposed of in boreholes with a diameter of 0.9 m. The heat
production of the HAW will be about 900 W per glass block ten years after the fuel
elements have been taken from the reactor core. Their borehole diameter will
be 0.4 m. The KPE planning presented here does not include heat calculations for
this type of waste. Parameter studies of Entwicklungs-Gemeinschaft Tieflagerung
(EGT) and BGR show however, that NEZ’s heat-producing waste over a 50-year
period, as characterized above, can only be disposed of in the envisaged disposal
IAEA-SM-243/48 307
— salt transport-------waste transport^ waste transfer port
sail transfer pant ■A disposal
—c» fresh air —► exhaust a ir HCL main mine fan WA ventilated sections ¿ !-r collection drift far
exhaust & r CS disposal floor SS sail transport level USt undercut drift V f tumble-cbwn
technique SI remote stocking
technique AT top loading chamber ВТ vertical bcre hole
technique
FIG.10. Flow-diagram of waste disposal and salt transport.
area if vertical 300 m deep boreholes can be realized. Assuming for example the
following conditions for MAW (HAW):
maximum temperature at the rim of the borehole < 100°C (200°C);
borehole depth 300 m (300 m);
distance between the boreholes in the drift 13 m (50 m);
distance between two drifts 16 m (50 m);
about 20 (7) boreholes are needed per annum. They would cover an area of
roughly 4000 m2 (20 000 m2).
For these reasons, among others, PTB is planning engineered storage facilities
for HAW, which will allow the storing of the HAW — if necessary — for decades,
thus enabling the heat production to be adjusted to the capacity of the dome and
the drilling technology available.
Figure 10 gives a schematic view of the disposal techniques described above
and shows the waste and salt transport scheme and the ventilation sections.
4.4. Backfilling and sealing of the mine
In the present concept the excavated salt is used to backfill and seal the
disposal chambers and boreholes. After use a whole disposal field is additionally
sealed from the rest of the operating mine by horizontal dams consisting of
shaft ¡ ¡ shaft /
308 RÔTHEMEYER
concrete and clay [3]. The same technique is applied, when after the disposal
operations have been finished, the whole disposal floor has to be backfilled and
sealed. Because of the good quality of the impermeable shaft liner, the shaft is
backfilled with salt and sealed with a concrete plug.
5. TIME SCHEDULE
As has been mentioned before, deep drillings will take about two years. The
subsequent sinking of the two shafts using the freezing method will need roughly
six years. The driving of the mine can be completed after four to five years so that
the disposal of waste could be started twelve years after the beginning of the
deep drillings. This rough estimate has not taken into account the time needed
for the licensing procedure. Exact data can hardly be given as there is no
experience with a licensing process according to § 9 b of the German Atomic Law.
Taking into account these imponderables and the beginning of deep drillings late
this year, the mine should operate in the first half of the nineties.
REFERENCES
[1] KÜHN, K., RANDL, R.-P., ROTHEMEYER, H., “The Federal Republic of Germany’s programme for the disposal of radioactive waste” these Proceedings, SM-243/95.
[2] LEVI, H.W., “Project “Safety-Studies Entsorgung” in the Federal Republic of Germany”, these Proceedings, SM-243/17.
[3] GRÜBLER, G., Der Baü von Dâmmen in Salzbergbau, Kali und Steinsalz, April 1976.
DISCUSSION
J.K. WIERZCHON: For transport on the surface and underground do
you plan to use only standard equipment such as fork-lift trucks or are you
thinking of other kinds of adapted or specially-designed vehicles?
H. ROTHEMEYER: The waste containers are to be transported by rail to
the shaft down to the disposal level. The whole transport unit (wagon and
containers) is then loaded on a truck and transported to the disposal area.
Depending on the disposal technique, the waste barrels are then brought to their
final disposal position by means of motorized shovels or cranes.
J.K. WIERZCHON: Will the canisters and drums with medium- and high-
level wastes be taken to the shaft in radiation-protection casks?
H. ROTHEMEYER: Yes, they will be transported in radiation-protection
casks.
IAEA-SM-243/48 309
G. ROCHLIN : I notice that you are drilling at the edge of the dome to the
north-west and the south-east. I wonder if, in your extensive testing and drilling
programme, you are making any studies to determine the extent of the dome
towards the north-east, under the Elbe and to the west of the village of Lenzen
in the German Democratic Republic. Or is it believed that lack of data or political
control over a potentially large part of the formation is of little importance for
safe operation and storage?
H. ROTHEMEYER: The extent of the dome towards the north-east is
generally known. It is believed that the data obtained from the site investigation
programme will be sufficient to guarantee safe operation of the repository,
especially taking into account the safety zone of at least two kilometres between
the river Elbe and the disposal area.
J. HAMSTRA: You allow a dose rate of 200 mrem/h for barrels with low-
level waste. Is the experience in the Asse mine such that you really can allow that
upper limit for emplacement in big rooms without exceeding the exposure
limits for the disposal personnel?
H. ROTHEMEYER: Present transport regulations require dose rates of not
more than 200 mrem/h at the surface and 10 mrem/h at a distance of 1 m from
the container. The transport system and the handling equipment for the Gorleben
repository afford a guarantee that the dose limits laid down by the Radiation
Protection Regulations will not be exceeded.
A.G. JACOBI: Is there not a contradiction in the fact that studies on the
storage of high-level radioactive wastes in connection with your site-independent
project are more or less completed even though little or no experience on such
storage is available, in contrast to the situation with low- and medium-activity
waste, which has been the subject of a large number of experiments at the Asse
salt mine?
H. ROTHEMEYER: Our conceptual design includes the disposal of high-
level waste into boreholes. Before this technique can be applied on an industrial
scale further research and development work is needed, especially on the rock
mechanical behaviour of the repository area. These problems are being tackled
under broad research and development programmes in the Federal Republic of
Germany and elsewhere.
K. KÜHN {Chairman): Mr. Jacobi’s comment is based on an erroneous
assumption. In fact, a good deal of information is available regarding disposal of
high-level waste, namely from laboratory investigations, in situ tests and computer
calculations. What is still lacking is the experience of a real test disposal operation
with high-level waste. The next five-year programme includes such a disposal
operation in the Asse salt mine.
R. KÔSTER: I should like to make a comment on the research and
development work which has been done on the disposal of the upper medium
active waste category. At the Kemforschungszentrum Karlsruhe calculations
310 RÔTHEMEYER
were made on the temperature profiles for the storage of this waste category.
Work was also performed on waste characterization (determination of nuclide
contents). More particularly, feed filtering data show that owing to the main
106Ru/106Rh component in (З/7 nuclides, there is a relatively quick decay in heat
production per barrel so that after, say, five years of cooling time, no temperature
problems will any longer arise.
M. KOMURKA: In the paper you say that retrievability was not considered
for safety reasons. Could you please explain what those safety reasons are?
If the storage of spent fuel was not considered for the Gorleben site, what kind
of retrievability were you thinking about?
H. RÓTHEMEYER: Our safety concept is based on the assumption that
before the waste is disposed of in an industrially operated repository sufficient
experience is needed to guarantee safe isolation of the waste without an interim
period, when retrievability might be considered. Backfilling and sealing immediately
after filling the rooms, boreholes and whole disposal areas is considered to
guarantee optimal isolation of the waste, especially under accident conditions.
As for your second question, we have in mind solid and solidified waste from the
reprocessing plant.
IAEA-SM-243/162
STUDIES ON THE OPTIMAL DISPOSAL OF RADIOACTIVE WASTES WITH SPECIAL ATTENTION TO THE THERMAL INFLUENCE ON THE SURROUNDING SALT BED AND TO ECONOMIC ASPECTS
A.S. KUNSTMAN, K.M. URBAÑCZYK,
J.K. WIERZCHOÑ
CHEMKOP,
Cracow
J. ^ASZKIEWICZ
Energoprojekt,
Warsaw,
Poland
Abstract
STUDIES ON THE OPTIMAL DISPOSAL OF RADIOACTIVE WASTES WITH SPECIAL ATTENTION TO THE THERMAL INFLUENCE ON THE SURROUNDING SALT BED AND TO ECONOMIC ASPECTS.
This paper presents the method commonly applied in Poland of forecasting the thermal character of an underground high-level radioactive waste repository. It is used for optimizing the mine excavation configuration as well as the order of storage. This method is shown with an example of the research results for the central radioactive waste repository. A short description of the designed repository as well as a comparison of expenditure for underground and surface repositories is given. The method shown in this paper of forecasting the temperature rise in the repository is based on the superposition of the analytical integral solutions of the heat conductivity equation for a single canister — the heat source. The detailed computer tabulation of these solutions enables forecasting the possibility of fast temperature increase at any repository point and at any time, taking into consideration localization and storage time for each of the thousands of stored canisters individually. In the repository design phase this allows the speedy investigation of various variants in the configuration of the placement and transportation corridors as well as the choice of a variant ensuring the best utilization of the repository area and also suitable working conditions for both mining and placement crews.
1. INTRODUCTION
According to the Polish programme of nuclear energetics development,
the first nuclear power plant is to be put into operation in 1984. Future planning
anticipates building of condensation, thermal-electric large-sized power plants
311
312 KUNSTMAN et al.
TABLE I. RELATIVE VALUES FOR THE TOTAL INVESTMENT COSTS
OF DIFFERENT REPOSITORIES
(100 = expenses for surface repository up to 1989) .
Variant Repository typeto 1989
Total investment costs to 2000 to 2020
Surface 100 176 1573A
Underground 489 521 711
Surface to year 2000
later 100 626 990
Вunderground
Onlyunderground 489 558 859
for heating purposes. Studies and analyses of the optimal development of
nuclear energetics in Poland take into account world experience to date, and
particularly safety and environmental protection aspects. One of the more
important questions considered is the problem of neutralizing and storing the
radioactive wastes which arise from nuclear power plants. Also the problem of
storing high-level wastes (HLW) from the fuel cycle is considered, since in future
the development of nuclear energetics will have to be combined with fuel
reprocessing and regeneration. To show the scale of the problem, 23 000 MW -
of power are foreseen to be installed in nuclear plants by the year 2000 and
power doubling would occur every ten years up to 2 020 .
2. POSSIBLE SOLUTIONS OF STORAGE PROBLEMS AND THEJR
ECONOMIC ANALYSIS
Radioactive wastes will be processed and neutralized. Solidification of
liquid wastes in bitumen and plastics, compression of solid ones and combustion
of some of them are planned. Low- and medium-level wastes (LLW and MLW)
will be packed into 200 dm3 barrels, and the HLW from the fuel cycle will be
solidified within a glass matrix in 50 dm3 canisters of diameter 30 cm. Two
possible variants, i.e. surface and underground waste storage, are examined in
detail. Storing on the sea bed has not been considered.
IAEA-SM-243/162 313
For the central surface repository the solution proposed is to use modular
reinforced concrete chambers or trenches placed in isolating clay, covered with
prefabricated plates isolated from precipitations and covered with soil.
The repository site will be suitably drained and will be successively enlarged.
The second variant is a central underground repository in the salt bed,
specially designed for this purpose with appropriate mining methods. Such a
repository would accept all kinds of radioactive wastes - with the LLW and the
MLW barrels located in large underground chambers while columns of a few
HLW canisters would be placed in shallow holes bored from the placement corridors.
Interesting results are obtained from an economic comparison of both
types of repository. This analysis has been made for two different solutions.
In variant A it has been assumed that no HLW from the fuel cycle is stored
while in variant В it is assumed that such a necessity will arise as from the
year 2000. In Table I the relative size of the total investment costs of
repositories are shown (assuming an index value of 100 for the expenses for
surface repository building up to 1989).
As can be seen from Table I, Polish conditions also corroborate that surface
storage may be considered only for minuscule power levels in nuclear plants.
For the wider range of nuclear power industry development programmes, and
in particular anticipating the storage of fuel cycle wastes, the building of a
repository in the salt bed — despite higher expenditure in the early stages -
is the more suitable solution from both the technical and the economic standpoints.
3. SHORT DESCRIPTION OF THE DESIGNED SALT BED REPOSITORY
The repository will be located in the Baltic seashore region, in the
Zechstein deposit of rock salt. The salt roof lies approximately at a depth of
740 m on average. The mean salt bed thickness in the repository region is
about 200 m. The overburden is composed of anhydrites and dolomite limestones
from Zechstein and argillaceous gritty formation from Mesozoic and Cenozoic
in which water-bearing horizons can be expected. For this reason a 120-m-thick
protective salt layer above the repository is planned.
The life of the central waste repository is put at about 50 years. All the
underground excavations will be designed and built with special attention to the
optimization and the safety of storage. Underground enlargement of the
repository will follow in stages, with two crews working below the ground:
a placement crew and a mining crew which will prepare new corridors and
chambers.
The sinking of three pit shafts using special methods is foreseen. Repository
corridors will be made using mechanical mining and the chambers will be blasted.
The repository ventilation will be divided into two air circulation circuits:
314 KUNSTMAN et al.
“pure” — through mining excavations, and “polluted” — through the
excavations already filled up. An appropriate time schedule ensures there will
be no overlap between the mining work period and the placement one.
The repository area planned for the HLW will be divided into rectangular
3.5 ha fields. Each field will contain a few parallel placement corridors, each
150-200 m long, joined at both ends by means of transportation corridors.
After filling up the whole field with canisters, which will take a few
months, the field will be isolated with bulkheads from the rest of the repository
so as not to overload the ventilation system. Possible backfilling of the already
loaded corridors with the salt output before their bulkheading is also under
consideration. The distances between the placement corridors as well as the
distances between the boreholes for canisters in the corridor should be optimized
in the course of the planned investigation of the temperature distribution
in this past of the repository.
An outline of the repository design and its features can be seen in Fig.l.
4. TEMPERATURE DISTRIBUTION PROGNOSIS IN THE
HLW REPOSITORY
4.1. Purpose of the forecast
The main purpose of thermal calculations was to forecast the temperature
increase in placement as well as in transportation corridors. This forecast
serves in the repository design phase as a basis of optimization of borehole
location and order of storage. The task of the optimization procedure is to
ensure the least possible temperature increase in corridors during their usage
by the crew.
The method of thermal calculations used to obtain such prognosis should
therefore fulfil the following requirements:
A short time only should be needed for computer calculations and
simplicity in studying different variants of repository configuration is
desirable.
Account must be taken of the location and time of storage of each individual
waste canister.
A secondary task was to estimate the temperature of the canister itself
after storage and to estimate the thermal influence of the repository on the
environment, points usually stressed in the literature [1].
IAEA-SM-243/162 315
I ¿ IW -s h a ll Mming Exhaust
Ж Man and Material shaft Air Jntake
IML W, hl W -shaft 1 = Placement Exhaust
1*00
HLWi I
НШ
Ж. Men and Material' Mom Corridors (2 *) Placement Main Corridor
шип. lu*I□¡s3iS|S S 2 2щ щ и з е е е е ;0 ® 2 ¡ E ¡a и 2¡ 0 ESI Ш1a S S 3 E E 3 3 '□ 00 E ES 2¡ a Ш s□ □¡аИЕЗаЕ;ЗИ □□□□ЕЕЕЗНиышшше'еие
-~Н1>V Area
Repository layout
, • — ÎIIH
Concrete plug hl f t canisters
LLiv- chamber
FIG.l. Repository outline. Dimensions in metres.
4.2. Theoretical model o f thermal calculations
The repository design considered by the authors involves the placement of
tens of thousands of HLW canisters, each of which - as mentioned above —
should be considered individually. For this reason it was not possible in practice
to use mathematical simulation methods of heat distribution with finite-element
or finite-difference approximation. The only real possibility left was to use
superposition of analytical solutions for individual canisters.
316 KUNSTMAN et al.
the rock surrounding the stored canisters is entirely homogeneous and
forms a half-space cut from above by a surface of constant temperature;
the stored canister is replaced in the calculation scheme by a segment,
neglecting its thickness.
For the above assumptions the following analytical solution of the
differential equation of the heat conductivity can be found [2]:
Such a method required the introduction o f some additional assumptions:
t
where:
erf (x) : du
T is the temperature rise
r is the distance from the container axis
z is the vertical ordinate
t is time
di is the canister top ordinate
d2 is the canister bottom ordinate
Q(r) is the thermal power of canister (time-dependent)
1 / Гa = — v — , where :
2 pc
IAEA-SM-243/162 317
X is the heat conductivity coefficient of salt rock
p is the density of salt rock
с is the specific heat of salt rock
It should be noted while estimating the accuracy of formula (1) that the
only important simplification is the assumption of the rock homogeneity
(neglect of the corridors and the ventilation influence) because the others — as
follows from our calculations - have no practical influence on the results.
4.3. Numerical tabulation of formula (1)
The complicated analytical form of formula (1) requires computer
tabulation. In such calculations the only problem consists in the accuracy of
numerical integration since the integrated functions vary greatly. The authors
have examined a series of numerical integration methods, from the simple
trapezia and Simpson methods to the more sophisticated Newton-Coates
integration method and finally decided to use the third-order Romberg method [3].
It was important to obtain high accuracy in a short time for the computer
calculations. Therefore integration with the step number depending on the
The tables have been obtained for a separate 2.8-m-long source
located in a hole bored from the corridor at a depth of 2—4.8 m from the
corridor floor. It has been assumed that the half-life period of the HLW
stored in the source would be 30 years on the average and the source power
on the storing moment would be 4 kW. Tables have been prepared for z
at the corridor floor. The range of distances r varies from 0 to 200 m,
and the time t from 0 to 100 years (Fig. 2). The time range was derived
from the repository working period of interest to the designers. Cal
culations have shown that in the period up to 100 years the thermal influence
of a separate source at a distance of 200 m is not greater than 10-5 °C, the
tabulation of larger distances was therefore unnecessary. The table included
over 10 00 0 values calculated for different times and distances with steps of
1 year and 2 metres respectively.
4.4. Attempt to parametrize the T(r,t)
In order to obtain a simpler analytical form of formula ( 1 ) the authors
fitted the exponential-polynomial functions to the values from the tabulation
of the function T(r,t). The calculations were made by means of the least-squares
method for six different types of function. The best results were obtained for
the following function:
318 KUNSTMAN et al.
FIG.2. Temperature rise resulting from the single 4 kW source as a function of time and
distance. The curves have been obtained from formula (1 j for the level z = 2 m above
the source top, \ = 1.33 cal/(m-s-°Cj; С = 0.22 cal/(g-°C); p = 2.17 t/m3'. The number of
steps providing the needed accuracy of the numerical integration is given for each domain.
T (r,t) = exp(Í Èi=0 j=0 '
(2)
With a third-order polynomial about 88% of the expected variability is
reproduced, and with an eighth-order polynomial as much as 99.3%.
This adjustment still had, however, in some small areas residua of up to
20% (in relation to the table data). Since the forecast of the temperature distribution
in corridors requires superposition of temperature contributions resulting from
thousands of sources, such deviations could accumulate and result in an
unnecessary mistake in the forecast.
Therefore it was decided to use the tables directly, from which the
program suitably interpolated the values needed.
All the programs mentioned in this paper, i.e. the program TANT which
does the tables of formula (1), the program F ANT - which does table
parametrization, and the program MANT — which makes the appropriate
IAEA-SM-243/162 319
forecast, using those tables, were elaborated by A. Kunstmann and K. Urbañczyk
in the computer language FORTRAN IV-1900 (the ICL version) and used on
the Polish computer ODRA-1305.
4.5. Characteristics o f the variants examined
Thermal prognoses have been performed for the design of the HLW
repository section which — in the course of 30 years — should store almost
100000 canisters each of 1 kW (at the time of initial storage). It has been
assumed that one borehole would contain four canisters, one above the other.
The whole region considered was divided into a few tens of storage fields with
about 1200 canisters in each, which amounts to 300 sources of power of
4 kW per field.
Different configurations of placement corridors inside the field as well ■
as different distances between placement boreholes in the corridor have been
considered. Different interfield pillars and different storing orders of individual
fields were also examined.
From the economic point of view it is better to store canisters as close as
possible in the placement corridors, whilst the safety and ventilation precautions
favour large-distance storing. One of the tasks of the forecast was therefore to
find the optimal canister distance.
To calculate this the program has to know the location and storage time of
each individual heat source. In order to avoid feeding into the computer tens
of thousands of such numbers — different for each storage variant — the program
MANT was prepared in such a way that for each elementary storing field this
information is read only for the first and the last field source while the rest are
suitably interpolated. Calculation of the forecast temperature rise in a given
time at one repository point takes from 0.5—0.9 s on the computer ODRA-1305.
4.6. Results o f the forecast
With the help of the program MANT the prognosis of temperature rise
in the repository for different storage variants has been made. Due to lack of
space we can in this paper give only some of the results of the variant considered
as optimal. In this variant an elementary storing field consisted of 10 placement
corridors, each 165 cm long with 30 placement boreholes. The distance between
adjacent boreholes was 5 m, the pillars between the corridors were 15 m wide,
the corridor width was 4 m, the distance of transportation corridors from the
extreme placement borehole was 10 m (Fig.l).
The forecast was made for a 100-year period. All the results given below
and in Figs 3 and 4 refer to the temperature rise, so that these values are to be
added to the ambient temperature in the repository before disposal.
320 KUNSTMAN et al.
¡ÿears]
FIG.3. Diagram of time-dependent temperature rises at the choosen HLW repository points
---------------- — above the central canister of field 1.
---------------- ---- above the central canister of field 40.
---------------- — above the first stored canister.
------ --------- - above the last stored canister of field 1.
------ --------- - above the last stored canister of field 40.
------ --------- - in the transportation corridor o f field 1.
------ • ------ - in the transportation corridor of field 40.
The field numbers are the same as in Fig.l. Field 40 is the central field o f the repository
fragment considered and field 1 is its peripheral one.
The main conclusion to be drawn from this forecast is that although the
temperature rises in the placement corridor will be great — up to 50°C — this
will occur relatively late — only after about 35 years. A year after storing, the
temperature rise of the placement corridor floor will be a little greater than 10°C,
after two years 18°C, and after five years this rise will be almost 30°C (Fig.3).
Here one can draw the conclusion that not later than one year after the
end of storing in the given field this should be bulkheaded from the rest of the
repository and switched off from the normal ventilation circulation to avoid
its thermal overloading.
The temperatures in transportation corridors rise much more slowly
(Figs 3, 4). After 5 years the rise is less then 2°C, after 10 years 5°C, after
25 years 13°C. The reason for this is a 10-m-wide pillar between transportation
corridors of adjacent fields which was added as a result of the investigation of
other variants. Such a procedure removed the transportation corridor from
the thermal influence of the adjacent field.
IAEA-SM-243/162 321
FIG. 4. Diagrams of temperature rises at the cross-section through field 59 versus the main
corridors for different times
---- ------- -----------after 3 years
------------------------ after 10 years
------------------------ after 40 years
— ------ — ■ — - after 70 years
------------------- ---- after 100 years
The time is counted from the storing moment of field 59.
The investigations on the order of storage by fields led to the choice of
the variant shown in Fig. 1 where the numbers in the fields correspond to this
order. By choosing such an order one avoids storing canisters in the neighbour
hood of fields filled so long before that a stronger thermal influence could
arise. The thermal forecast showed that with such a time schedule of filling
the fields the influence of the already filled regions on the current storing place
is almost unmeasurable. In other words, the placement front moves faster than
the front of the explicit temperature rise. This fact ensures full freedom in the
execution of mining and placement work. Another important feature of the
chosen storing order is the simplicity of bulkheading the already filled fields
from the rest of the repository. After field closing a small directed air flow
through the old transportation corridors will be maintained in order to evacuate
possible gaseous impurities which would come from the bulkheaded placement
corridors.
322 KUNSTMAN et al.
In the chosen variant the density of the initial waste power is in the field
centre 485 kW/ha, and for the HLW repository on average 345 kW/ha (with
inside transportation corridors and their pillars already added). These values
are given only to enable comparison; it would be wrong, however, to consider
merely the power density without regarding the storing schedule and the local
repository structure.
The authors claim that the accepted repository configuration is quasi-
optimal. More widely-spaced storing is pointless as this makes costs unnecessarily
higher. A variant of closer storing would make sense only if the number of
canisters in each placement borehole were increased from 4 to 5. Then,
however, the temperature rises would be correspondingly higher, e.g. the
temperature rise in the placement corridor would reach 63°C, which seems
too high.
In addition, the authors estimated the temperature at the canister edge
after storing in the placement borehole. This was done by means of the
stationary model based on formulas given in Ref.[4]. After adding to the result
for that model (for a single source) the maximal influence of adjacent sources
(calculated by the programme MANT), the temperature rise of the edge of the
4 kW source with reference to the bed temperature has been estimated at
about 165°C. A temperature of the order of 200°C is admissible for the canister.
4.7. Additional remarks
The results presented in section 4 are only the first findings of the work
begun a year ago and were obtained by a small group in a few months. Further
work will consider the influence of the ventilation system as well as of the
corridor shape.
5. CONCLUSIONS
(1) The central radioactive waste repository in a salt bed is feasible and
economic in Poland. It will doubtless be necessary if HLW disposal is considered.
(2) The thermal prognosis method shown in this paper is a fast and efficient
instrument for optimizing the HLW repository configuration in the design phase.
(3) The most important feature which characterizes the temperature
distribution in the repository is its extremely slow rise with time. The temperatures
can reach high values (a few tens of degrees) but this happens only after some
tens of years.
(4) The thermal influence of an underground repository on the salt bed
overburden (taking also into consideration the possible water horizons) can
emerge only after a few centuries and will be of small significance. Therefore
an environmental protection problem does not arise.
IAEA-SM-243/162 323
(5) Computer simulation methods of heat flow in the repository using the
approximation with finite elements or differences are worth putting into use
only as more precise local statements of the thermal forecast for the chosen
repository sections, e.g. for considering the corridor shape and the influence of
ventilation or for examining the immediate canister surroundings (air slot,
concrete plug). Boundary conditions for these sections of the repository can
be taken from the analytic-superpositional model described above.
(6 ) Due to the thorough investigation of a number of storage variants it
was possible to choose a solution which assured the full thermal safety of the
placement and mining work as well as a high degree of repository space
utilization and close storage in the corridors at the same time.
REFERENCES
[1] United States Working Draft on Repository Physical Descriptions in Salt Formation, United States Department of Energy, May 1978.
[2] CARSLAW, H.S., JAEGER, J.C., Conduction of Heat in Solids, Oxford Univ. Press,London, New York (1959).
[3] RALSTON, A., A First Course in Numerical Analysis, Me Graw-Hill, New York (1965).[4] WHEELER, B.R., et al., “Storage of radioactive solids in underground facilities —
current ICPP practices and future concepts”, Disposal of Radioactive Wastes into theGround, (Proc. Int. Symp. Vienna, 1967), IAEA, Vienna (1967) 421.
DISCUSSION
J.A. ANGELO: With respect to possible storage of spent fuel or high-level
waste, have any of your thermal calculations dealt with the case of phase change,
i.e. the melting of some surrounding geologic materials?
K.M. URBAÑCZYK: In our model the highest temperature on the canister
surface will not exceed 200° C.
J.A. ANGELO: If spent fuel or high-level wastes are eventually to be
disposed of within Polish territory, are you optimistic about a favourable
public response?
K.M. URBAÑCZYK: We do not think problems will arise in this connection.
J. HAMSTRA: What isolation period do you expect to obtain in the long
term with the 120-m-thick protective salt layer above the repository, taking
into account expansion movement of the salt bed due to the thermal load of
the high-level waste?
K.M. URBAÑCZYK: We believe that this protective layer is sufficient to
ensure permanent isolation.
ИССЛЕДОВАНИЕ РАДИАЦИОННОЙ СТОЙКОСТИ
ПРИРОДНОЙ КАМ ЕН Н О Й СОЛИ
В.И. СПИЦЫН, Л.И. БАРСОВА, С.А. КАБАКЧИ,
И.И. ЗЯЗЮЛЯ, И.Е. ЛЕБЕДЕВА
Государственный комитет по использованию
атомной энергии СССР,
Москва,
Союз Советских Социалистических Республик
Abstract- Аннотация
INVESTIGATION OF THE RADIATION STABILITY OF NATURAL ROCK SALT.The paper describes a study of the physical and chemical processes occurring in rock
salt samples taken from various deposits when subjected to gamma radiation in a wide range of doses. The nature and kinetics of the build-up of electron and hole centres are examined as a function of impurity content, absorbed dose and temperature. The formation of point F-, M- and V-centres is demonstrated, and the processes of thermal and radiation-induced coagulation into colloidal sodium with gaseous chlorine are discussed. The dose absorbed by a crystal can be determined from the number of F-centres that have formed, and the effectiveness of the liberation of gaseous products, mainly molecular chlorine, depends on the impurity content. With the dissolution of irradiated samples of rock salt, molecular hydrogen and hydrogen peroxide are formed. There is a linear relationship between the quantity of hydrogen formed and the concentration of F-centres in the sample. A technique for determining the energy stored by the crystal from the quantity of hydrogen liberated is proposed. A study is made of the microhardness and brittleness of rock salt samples from various deposits irradiated with a wide range of doses (up to 8000 Mrad). It is shown that the kinetics of microhardness variation and the maximum radiation hardening of crystals depend on the accumulated (‘biographical’) defects and impurities.
ИССЛЕДОВАНИЕ РАДИАЦИОННОЙ СТОЙКОСТИ ПРИРОДНОЙ КАМЕННОЙ СОЛИ.В настоящей работе исследованы физико-химические процессы, протекающие в образцах
каменной соли различных месторождений при воздействии 7- излучения в широком интервале доз. Изучены природа и кинетика накопления электронных и дырочных центров в зависимости от содержания примесей, поглощенной дозы и температуры. Показано образование точечных F-, М-, V-центров и изучены процессы их термической и радиационной коагуляций вплоть до коллоидального натрия и газообразного хлора. По количеству образовавшихся F-центров можно судить о поглощенной кристаллом дозе. Эффективность выделения газообразных продуктов, главным образом молекулярного хлора, зависит от содержания примесей. При растворении облученных образцов каменной соли образуются молекулярный водород и перекись водорода. Существует линейная связь между количеством образующегося водорода и концентрацией F-центров в образце. Предложен способ определения запасенной кристаллом энергии по количеству выделившегося водорода. Изучено изменение микротвердости и хрупкости образцов каменной соли различных месторождений, облученных в широком интервале доз (до 8000 Мрад). Показано, что кинетика изменения микротвердости и максимальное радиационное упрочнение кристаллов зависит от накопленных ("биографических”) дефектов и примесей.
IAEA-SM-243/109
325
326 С ПИЦЫ Н и др.
Захоронение твердых радиоактивных отходов в отработанных пластах каменной
соли рассматривается в настоящее время как перспективный метод удаления отходов с
поверхности Земли, позволяющий предотвратить заражение среды обитания [1,2] .
Решению вопроса о возможности захоронения отходов в конкретном месторожде
нии должно предшествовать исследование изменения физико-химических свойств ве
ществ, слагающих породу, в поле ионизирующего излучения. Одним из важнейших
свойств породы является ее радиационная стойкость, т.е. способность не изменять под
действием излучения те физико-химические характеристики, которыми обеспечивается
надежность хранилища в течение длительного времени (сотни лет).
Кроме исследования самой радиационной стойкости имеется настоятельная необ
ходимость разработки методологических аспектов прогнозирования изменения пара
метров хранилища с течением времени при заданных условиях его эксплуатации, а так
же прогнозирования последствий возможных аварий.
В решении этих проблем существенную помощь может оказать информация, име
ющаяся в мировой литературе, о поведении щелочно-галоидных соединений при воздей
ствии ионизирующего излучения (например, [3, 4] ). Однако, практически все данные
получены с искусственно синтезированными соединениями. В то же время хорошо из
вестно, какую важную роль играют ’’биографические” дефекты и примеси в создании
радиационно-химических нарушений в твердых телах. Поэтому для использования име
ющихся литературных данных необходимо определить, насколько основные закономер
ности радиолиза солей данного месторождения соответствуют закономерностям для
синтетических кристаллов. Кроме того, важной задачей является разработка простых
экспериментальных методов количественного определения радиационных нагрузок на
вещества породы предполагаемого хранилища.
В настоящей работе исследованы физико-химические процессы, протекающие в
образцах природной каменной соли при воздействии у-излучения в широком интервале
доз и температур, т.е. в модельных условиях, имитирующих подземное хранилище.
Основная часть работы проведена с образцами природной каменной соли, практи
чески не содержащими примесей (в пределах чувствительности лазерного анализатора) .
Другой тип изученных образцов природной каменной соли содержал, согласно рентгено
флюоресцентному анализу, большие количества примесей (К, Mg, Са, Si, Al, Fe). В
дальнейшем мы будем обозначать эти типы кристаллов соответственно как I и II типы.
1. ЦЕНТРЫ ОКРАСКИ В ОБЛУЧЕННЫХ КРИСТАЛЛАХ КАМЕННОЙ СОЛИ
Хорошо известно, что радиационные нарушения в щелочно-галлоидных солях,
такие, как, например, образование газообразного галогена, коллоидного металла, из
менение твердости, хрупкости и т.д., определяются первичными радиационно-хими-
ческими дефектами, условиями их образования и взаимных превращений. Поэтому
в первую очередь были изучены природа и кинетика накопления первичных радиаци
онно-химических дефектов (центров окраски) в образцах каменной соли.
IAEA-SM-243/109 327
ДОЗА (Мрад)
Рис. 1. Кинетические кривые накопления центров окраски в природных образцах каменной соли
I типа, подвергнутых y-облучению при 300К: кривая 1 - концентрация F-центров (т\), кривая 2 -
концентрация М-центров (пм1’ кривая 3 - концентрация V-центров (fty)-
Методами оптической спектроскопии и электронного парамагнитного резонан
са в у -облученных при ЗООК кристаллах NaCl были идентифицированы электронные
F-центры (электроны, захваченные в одиночных анионных вакансиях), дырочные
V-центры, представляющие собой дырку, захваченную междоузельным ионом галоида,
М-центры (электроны, захваченные в двух соседних анионных вакансиях). При увели
чении дозы облучения до 1000 Мрад происходит агрегирование F-и М-центров и обра
зование коллоидного натрия. По спектрам оптического поглощения изучена кинетика
накопления центров окраски при ЗООК в зависимости от поглощенной дозы. Данные
представлены на рис. 1.
Из рисунка видно, что для всех центров окраски наблюдается двухстадийный
рост накопления. Для V- и F-центров перегиб на кривой накопления происходит
при ~ 1600Мрад, а для М-центров — при 10 Мрад. При 3300 Мрад насыщения на кри
вых накопления не достигается.
328 СПИЦЫ Н и др.
Максимально достигнутая концентрация F-центров в образцах каменной соли, не со
держащих примесей, составляла ~З Ю 18см_3, М-центров-- 41 016см"3. Если принять,
что максимальная концентрация дефектов в кристаллах может достигать 1 0 20- 10 11 см~э,
то полученные величины указывают на большую радиационную стойкость природных
кристаллов. Для образцов каменной соли, содержащих примеси, максимальная кон
центрация центров окраски почти на два порядка выше. По прямолинейному участку
кривой накопления центров окраски можно рассчитать их средние радиационно-хими
ческие выходы. В образцах природной каменной соли, не содержащих примесей, они
составляют для V-и F-центров 5-10-3 / 100эВ. В образцах, содержащих примеси, ради
ационно-химические выходы V-и F-центров достигают 10_1/100эВ. Поскольку на кри
вой накопления F-и V-центров в облученных образцах (кривые 1 и 2 на рис. 1) наблю
дается значительный прямолинейный участок (до ~ 1600Мрад), то по количеству обра
зовавшихся F- и VrueHTpoB можно судить о поглощенной дозе. В то же время скорость
накопления центров окраски может служить мерой радиационной стойкости каменной
соли.
Вследствие того, что радиационные отходы характеризуются весьма высоким
энерговыделением, в пластах каменной соли при захоронении могут создаваться по
вышенные температуры. Поэтому следующим этапом нашей работы было изучение
процессов термических превращений и коагуляции центров окраски. Методом опти
ческой спектроскопии было показано, что при нагревании 7-облученных при 300К об
разцов каменной соли F-центры агрегируются в М-центры (при 350-400К),при450-550К
— в центры коллоидального натрия, которыеисчезаютпридальнейшем подъеме темпера
туры. Кажущаяся энергия активации гибели F-центров мало зависит от поглощенной
дозы и составляет 0,4эВ; энергия активации образования М-центров — 0,3 эВ; а энер
гия активации гибели полосы коллоидального натрия - 1,1 эВ. При отжиге облученных
при 300К образцов наблюдается два пика термолюминесценции — при 333К и 543К,
которые можно объяснить рекомбинационными процессами с участием М-центров
(низкотемпературный пик) и центров коллоидального металла (высокотемператур
ный пик). Энергия активации разгорания люминесценции в области низкотемпера
турного пика составляет 0,8эВ и не зависит от поглощенной дозы; в области высо
котемпературного пика энергия активации уменьшается при повышении дозы облу
чения от 1,7 до 1,0 эВ и близка к соответствующей энергии активации гибели центров
коллоидального натрия. Интенсивность низкотемпературного пика термолюминес
ценции растет с увеличением поглощенной дозы до 1-2Мрад и затем снижается до ну
ля при ~ 20Мрад. Интенсивность высокотемпературного пика растет вплоть до доз
~ 2 0 0 Мрад,причем интенсивность пика остается неизменной при хранении образцов
в темноте в течение длительного времени; это характерно как для I, так и для II ти
пов кристаллов. Поэтому данные по интенсивности высокотемпературного пика тер
молюминесценции могут быть использованы для дозиметрических целей.
При высокотемпературном облучении образцов каменной соли (373-600К, до
за до бМрад) возникаютр- и V-центры; М-центры наблюдаются лишь при 300-373К,
при более высоких температурах они не образуются. Кинетические кривые накопле
ния V- и F-центров приведены на рис. 2. Из рисунка видно, что при температурах
IAEA-SM-243/109 329
п у (1 0 15см 3) a)
Рис.2. Кинетические кривые накопления V-центров (а) и F-центров (б/ в кристаллах каменной со
ли, облученных при повышенных температурах.
300 и 373К кривые накопления имеют почти прямолинейный характер. При более высо
ких температурах они выходят на плато уже при 1-2Мрад. Радиационно-химические выхо
ды образования F-центров при повышении температуры облучения снижаются от
Ы 0 _3/100эВ(373К) до ~ 10“5/100эВ (573К) . Более сложный характер носит изменение
радиационно-химических выходов V-центров с повышением температурыоблучения—при
473К и 543Кони выше, чем при ЗООК, но при дальнейшем повышении темепературы
уменьшаются. Низкотемпературный пик термолюминесценции наблюдается в образцах,
содержащих М-центры, высокотемпературный пик проявляется в образцах, облученных
не выше 540К.
Комплекс проведенных исследований центров окраски в природных образцах
каменной соли позволяет найти решение для количественного определения поглощен
ной дозы стенками хранилища во время эксплуатации. В качестве дозиметрического
контроля могут быть использованы кривые накопления F-центров и кривые измене
ния интенсивности высокотемпературного пика термолюминесценции от поглощен
ной дозы.
330 С ПИЦЫ Н и др.
Образование молекулярного хлора при облучении каменной соли является весь
ма неблагоприятным фактором при эксплуатации хранилища, поскольку хлор хими
чески активен и может вызывать протекание химических процессов, в частности, спо
собствовать коррозии. Кроме того, хлор высокотоксичен.
Образование молекулярного хлора происходит за счет взаимодействия так на
зываемых ’’электронных дырок” в агрегированных V-центрах. Из работ, посвящен
ных радиолизу синтетических кристаллов, известно, что количество образующегося
хлора при облучении зависит от целого ряда параметров — дозы, температуры облуче
ния, а также от ’’биографии” кристаллов. Максимальные радиационно-химические
выходы хлора для синтетических кристаллов не превышают Ю -4 молекул/ 100эВ [5].
Нами была исследована зависимость выделения хлора от поглощенной дозы
при различных температурах облучения (максимальная поглощенная доза ~ 4Мрад).
Было найдено, что для образцов природной каменной соли I типа выход хлора практи
чески не зависит от температуры в интервале 300-673° С и составляет
0,045 ± 0,007 молекул/100эВ. В то же время для природных образцов, содержащих
примеси, радиационно-химический выход хлора оказался практически близким к
нулю. Эти эксперименты показывают, что использовать литературные данные по га-
зовыделению не представляется возможным, и для каждого конкретного хранилища
следует проводить свои модельные исследования.
2. ОБРАЗОВАНИЕ ГА ЗО О БРА ЗН О ГО ХЛОРА ПРИ7-ОБЛУЧЕНИИ КАМЕННОЙ СОЛИ
3. ИЗУЧЕНИЕ Г АЗОВЫДЕЛЕНИЯ ПРИ РАСТВОРЕНИИ ОБЛУЧЕННОЙ КАМЕННОЙ
СОЛИ
Известно, что при растворении облученных образцов синтетических кристаллов
каменной соли образуется водород. В случае аварии хранилища при прорыве в него
воды образующийся водород может составить с воздухом смесь, способную к взры
ву. Взрыв этой смеси может привести к выбросу радиоактивности на поверхность
Земли. Следовательно, для прогнозирования последствий аварии и разработки мер
по их ликвидации необходимо знать количественные закономерности выделения во
дорода.
При растворении облученных образцов каменной соли молекулярный водород
образуется за счет рекомбинации гидратированных электронов, возникших при разру
шении F-центров. При этом между концентрацией F-центров и количеством молекул
водорода имеется линейная связь: пр = п^где nF — концентрация F-центров (см-3) ,
nHj — концентрация молекул водорода (см-3) . При растворении облученных образцов
в воде, содержащей кислород, кроме Н2, образуется также и перекись водорода, кото
рая в дальнейшем разлагается с выделением кислорода.
IAEA-SM-243/109 331
пН ](10 7моль/г)
Рис. 3. ■ Зависимость между количеством F-центров в облученных образцах природной каменной соли
и количеством водорода, образующегося при их растворении.
Нами была определена зависимость количества выделившегося водорода от пог
лощенной дозы. Было показано, что количество водорода практически линейно воз
растает с дозой в интервале 150- 1600Мрад. Максимальное количество водорода, обра
зующегося при растворении 1г соли, облученной дозой 1600Мрад, составляет 6,5-10~7моль.
Это соответствует радиационно-химическому выходу G(H2) — 4-10*4 молекул/100 эВ.
Была обнаружена прямолинейная зависимость между количеством образовавшихся при
облучении F-центров и количеством выделившегося при растворении водорода (рис.З).
Существование такой симбатной зависимости позволяет по количеству F-центров надеж
но и быстро определить ожидаемое количество водорода в случае аварии. С другой сто
роны, из рисунков видно, что теоретическая зависимость, полученная при изучении син
тетических кристаллов, справедлива и для природных образцов. Это дает возможность
использовать известное из литературы [6 ] соотношение между количеством выделивше-£
гося водорода (МНт) и запасенной энергией Е: 2Ñ-M где N - число Авогадро. На-
ми получена величина запасенной энергии Е =< 3-1018 эВ/r для образцов каменной соли,
не содержащих примесей, при дозе 103 Мрад. Запасенная энергия является одной из ко
личественных характеристик радиационной стойкости кристалла. В частности, сопро
тивляемость облученного материала разрушению обратно пропорциональна количеству
запасенной энергии [ 7].
4. ИЗУЧЕНИЕ МИКРОТВЕРДОСТИ ОБЛУЧЕННОЙ ПРИРОДНОЙ КАМЕННОЙ СОЛИ
В щелочно-галоидных кристаллах взаимодействие радиационно-химических де
фектов с дислокациями внешне проявляется в изменении механических свойств кри
сталлов — твердости, хрупкости, предела текучести, прочности и т.д. Изменение меха
нических свойств породы хранилища может оказаться неблагоприятным и привести к
разрушению пластов, окружающих радиоактивные отходы.
332 С П И Ц Ы Н и др.
Рис. 4. Кривые изменения микротвердости в зависимости от поглощенной дозы для образцов камен
ной соли двух типов: а) образец каменной соли Iтипа, б) образец каменной соли Птцпа.
Нами было проведено изучение микротвердости и хрупкости образцов каменной
соли после '('-облучения в широком интервале поглощенных доз (до 8000Мрад). Опре
деление микротвердости проводилось методом вдавливания, который характеризует
сопротивляемость материала пластической деформации.
Из литературного материала можно сделать вывод, что изменение микротвердости
от поглощенной дозы имеет сложный характер: при малых дозах облучения наблюдает
ся разупрочнение кристаллов (дислокации становятся короче, подвижнее), а при даль
нейшем облучении сложные радиационные дефекты создают новые дислокации, движе
ние которых будет тормозиться другими кластерами дефектов — наступает упрочнение.
Нами показано, что кинетика изменения микротвердости от поглощенной дозы не
одинакова для природных образцов двух типов и зависит, особенно при малых дозах,
от ’’биографических” дефектов. Это подтверждается рис.4. Из рисунка следует,
что микротвердость образцов I типа (кривая 1) возрастает при увеличении дозы до
1600Мрад, а при более высоких поглощенных дозах остается неизменной. Максималь
ное упрочнение достигает ~20% по сравнению с необлученными образцами. При срав
нении с кривыми накопления центров окраски (рис. 1) оказывается, что до поглощенной
дозы ~ 1600Мрад кривые накопления F- и V-центров и кривые изменения микротвер
дости имеют прямолинейную зависимость от дозы, однако при более высоких дозах кон
центрация центров окраски продолжает возрастать (хотя и более медленно), а микротвер
дость не изменяется. Таким образом, для кристаллов I типа не наблюдается прямой кор
реляции между изменением микротвердости и концентрацией центров окраски. Возмож
но, что при облучении образуются некоторые структурные дефекты, с накоплением ко
торых связано образование дислокаций и центров окраски. Согласно рис.4 (кривая 2),
зависимость микротвердости от дозы для природных образцов II типа носит сложный
характер. При поглощенных дозах до ЮООМрад наблюдается уменьшение микротвер
дости, затем увеличение ее по сравнению с исходной величиной и далее кривая проходит
через максимум. Никакой корреляции с кривыми накопления центров окраски не на-
IAEA-SM-243/109 333
блюдается. Следовательно, при изучении изменения твердости пород хранилища под
облучением необходимо проводить модельные эксперименты для каждого конкретного
месторождения.
Изменение поверхностной прочности твердых тел, измеренной по микротвердос
ти, можно связать, согласно [8], с изменением их поверхностной энергии. Для двух
однотипных твердых тел существует зависимость:
Ох _ 3 f e
а2 \j Hi
где о — поверхностная энергия, а Н — микротвердость.
Увеличение поверхностной энергии кристаллов при облучении приводит к возник
новению внутренних напряжений в кристаллах, а это в свою очередь может вызвать рас
трескивание кристаллов, что несомненно нежелательно для условий хранения радиоак
тивных отходов. Для природных образцов каменной соли I и II типов максимальное из
менение поверхностной энергии при облучении составило ffi/a2 = 1,13. Такое изменение
поверхностной энергии, очевидно, невелико и не должно привести к образованию тре
щин. Действительно, экспериментально нами было показано, что облучение не приводит
к заметному возрастанию хрупкости образцов.
ЛИТЕРАТУРА
[ 1J Me CLAIN, W.C., ВОСН, A.L., NucL Technol. 24 (3) 1974.[ 2] АДАМ, X., КЕРНЕР, В., РИХТЕР, Д ., Труды научно-технической конференции т. 2 (Колобжег,
2-7 октября 1972) Варшава (1973) 11.[ 3) ВОРОБЬЕВ, А.А., Центры Окраски в Щелочно-галоидных Кристаллах, 2, изд. ТГУ, Томск (1968).[ 4] MARKHAM, J.J., F-centres in Alkali Halides, Ac. Press, N.J. (1966).[ 5] ШВАРЦ, K.K., КРИПСТАНСОН, Я.Ж., ТИЛИКС, Ю.Е., в сб. ’’Влияние облучения на неметалли
ческие кристаллы”, Минск (1968) 5.[ 6] ВОРОБЬЕВ, А. А., ЗАВАДОВСКАЯ, Е.К., КУЗЬМИНА, А.В., Запасенная Энергия в Щелочно-га-
лоидных Кристаллах, изд. ТГУ, Томск, (1969).[ 7] АНДРОНИКАШВИЛИ, Э., ПОЛИТОВ, Н., ВОРОЖЕЙКИНА, Л., ГЕТИЯ, М., Radiation Damage
in Solids, v. Ill (Proc. Symp. Venice, 1962) IAEA, Vienna (1963) 147.[ 8] КУЗНЕЦОВ, В.Д., Поверхностная Энергия Твердых Тел, ГИТТЛ, М. (1954).
IAEA-SM-243/97
SUMMARY OF UNITED STATES GEOLOGICAL SURVEY INVESTIGATIONS OF FLUID-ROCK-WASTE REACTIONS IN EVAPORITE ENVIRONMENTS UNDER REPOSITORY CONDITIONS
D.B. STEWART, B.F. JONES, E. ROEDDER
US Geological Survey, Reston, Virginia
R.W. POTTER, II
US Geological Survey, Menlo Park, California,
United States of America
Abstract
SUMMARY OF UNITED STATES GEOLOGICAL SURVEY INVESTIGATIONS OF FLUID- ROCK-WASTE REACTIONS IN EVAPORITE ENVIRONMENTS UNDER REPOSITORY CONDITIONS.
The interstitial and inclusion fluids contained in rock salt and anhydrite, though present in amounts less than 1 weight per cent, are chemically aggressive and may react with canisters or wastes. The three basic types of fluids are: (1) bitterns residual from saline mineral precipitation including later recrystallization reactions; (2) brines containing residual solutes from the formation of evaporite that have been extensively modified by reactions with contiguous carbonate or clastic rocks; and (3) re-solution brines resulting from secondary, dehydration of evaporite minerals or solution of saline minerals by undersaturated infiltrating waters. Fluid composition can indicate that meteoric flow systems have contacted evaporites or that fluids from evaporites have migrated into other formations. The movement of fluids trapped in fluid inclusions in salt from southeast New Mexico is most sensitive to ambient temperature and to inclusion size, although several other factors such as thermal gradient and vapour/liquid ratio are also important. There is no evidence of a threshold temperature for movement of inclusions. Empirical data are given for determining the amount of brine reaching the heat source if the temperature, approximate amount of total dissolved solids, and Ca:Mg ratio in the brine are known. SrCl2 and CsCl can reach high concentrations in saturated NaCl solutions and greatly depress the liquidus. The possibility that such fluids, if generated, could migrate from a high-level waste repository must be minimized because the fluid would contain its own radiogenic energy source in the first decades after repository closure, thus changing the thermal evolution of the repository from designed values.
1. INTRODUCTION
Evaporite deposits contain rock salt, anhydrite or gypsum, carbonate rocks
(limestone and dolomite), and subordinate shaly or sandy units. Both rock salt
and anhydrite are potential host rocks for repositories for high-level radioactive
wastes.
335
336 STEWART et al.
The United States Geological Survey (USGS) is a participant in the US
Department of Energy’s National Waste Terminal Storage Program, which from
1957 to 1978 was focused on rock salt for repository siting. USGS investigations
are both generic and site-specific and range in detail from regional to microscopic.
Investigations bearing on specific rock salt strata are under way in Utah and New
Mexico, and regional hydrological investigations related to rock salt for repository
use are being conducted in Mississippi, Louisiana, and New Mexico. Regional
geologic studies of rock salt in New York and Ohio are almost completed. These
investigations have yielded much new data on the process of dissolution of salt
bodies, the direction and amounts of flow of the fluids that react with rock salt,
and fundamental processes such as salt flowage and the origin of breccia pipes.
In this paper, we shall report only results of some geochemical studies
focused on the nature and behaviour of the fluids in rock salt. Although the
amounts of fluids in rock salt are, in general, limited, average contents possibly
being less than one weight per cent in strata of interest for repositories, these
fluids are chemically agressive, may react with canisters or wastes, and may
adversely affect the sorptive capacity and mechanical strength of rock salt [1].
1.1. Rock salt
1.2. Anhydrite
In 1978, the USGS began to evaluate the suitability of anhydrite as a medium
in which a radioactive waste repository might be constructed. Favourable factors
suggesting the use of anhydrite include low solubility and negative coefficient of
solubility to 235°C [2], high thermal conductivity, low thermal expansion [3],
a generally low content of fluid inclusions, and an apparently higher sorptive
capacity than in rock salt [4].
An inventory and summary of the geologic, physical, hydrologie, and
chemical characteristics of anhydrite occurrences in the conterminous United
States has been prepared [4]. All major evaporite deposits within the con
terminous United States have been evaluated in terms of the stratigraphie relation
ship of anhydrite units to the entire deposit, the thickness of the entire deposit,
and the relative abundance of anhydrite. In particular, beds of anhydrite thicker
than 30 metres and occurring at depths of 300 to 1500 metres have been
characterized from the literature in order to evaluate each bed for possible
further assessment as a repository horizon. Although most anhydrite-bearing
beds are impure, beds of apparently adequate thickness and purity have been
identified in several basins, and a USGS programme for their further assessment
is being planned.
IAEA-SM-243/97 337
We discuss below the origin of the compositions of the fluids contained in
evaporites, especially rock salt. We also report briefly on some new results on
the behaviour of inclusions of fluids within salt when subjected to a thermal
gradient and give data on the physical chemistry of bitterns (brines containing
significant quantities of Mg, Ca, or К chlorides), including new data for solutions
saturated with NaCl and CsCl or SrCl2. We also give a very brief review of new
instruments being developed by the USGS to measure the amount and distribution
of H20 in evaporite sections and to devise new methodology for dating minerals
in salt.
2.1. Origin of the brine compositions in evaporites
Present findings, essentially supporting those of Ref. [5], indicate that the
brine compositions associated with the principal evaporite sections of the United
States are of three fundamental types, reflecting the chemistry of marine and/or
meteoric waters and their interaction with the basin deposits. The three basic
types are:
( 1 ) bitterns residual to saline mineral precipitation, including later recristal-
lization reactions;
(2) sedimentary basin (‘oilfield’) brines which contain residual solutes from the
formation of evaporite and which have been extensively modified by reaction
with contiguous carbonate or clastic rocks; ánd
(3) re-solution brines resulting from the secondary dehydration of evaporite
minerals or solution of salines by undersaturated waters infiltrating the
evaporite strata. ^
Bittern compositions reflect the prior precipitation of carbonate, sulphate,
and halite from evaporated marine waters. Bitterns are typically saturated in
NaCl, high in К and Mg, low in sulphate, and variable in calcium. Their composi
tion depends on the saline mineral assemblage (including any Mg, К salts), the
precipitation history, and the nature of the normally small content of fine-grained
clastic material. Examples of such fluids have been reported from inclusions
from the Permian Wellington Formation of Kansas, the Silurian Salina Formation
of Michigan and the Permian Salado Formation of New Mexico [6 ]. Similar
compositions have been obtained for seepage in potash strata of the Salado
Formation of New Mexico [7, 8], from the Paradox basin of eastern Utah [9],
and the salt domes of Louisiana [10]. Presumably, the extreme composition for
these brines would be a very high magnesium chloride content resulting from
incongruent dissolution of K, Mg chlorides (carnallite or kainite), as described in
Ref.[ 11 ]. Recent analysis of a sample from the potash mine workings has also
revealed the most ‘evolved’ bittern composition from the Salado evaporites to date.
2. FLUID REACTIONS IN ROCK SALT
338 STEWART et al.
Sedimentary basin (‘oilfield’) brines are associated with the thick carbonate
or clastic strata peripheral to the evaporite basins and are most specifically
characterized by high calcium content, though sodium is most commonly still
the dominant cation. Sulphate concentration is normally very low. The Na-Ca
brines of the Mississippi embayment and deep Gulf Coast sediments (including
Florida), as well as the Appalachian basin, may be considered most typical. A
wide range of solute compositions have been found in the ‘oilfield’ brines of the
Permian-Pennsylvanian Paradox basin, Utah [9], whereas the Denovian Detroit
River Group of Michigan contains fluids heavily dominated by CaCl2 [12].
The origin of the high calcium content of brines in some areas is controversial,
but the explanation of widest application seems to be that it is the reaction of
highly magnesian solutions with calcium carbonate to form dolomite, and the
consequent return of Ca to solution, as discussed in Ref.[5], The trend may also
be attributable to, or further enhanced by, formation of Mg or К silicates, such
as chlorite, mica, or feldspar, with displacement of exchangeable Ca from
metastable clay phases. Sulphate is lost from solution by additional CaS04 precipitation or bacterial reduction. Further concentration of the alkaline earth
halides (and Br) may result from retention of solutes behind shale ultrafilters in
deep-basin subsurface fluid migration [13].
Re-solution brines are dominated by solutes derived from the dissolution
of evaporites by undersaturated waters, commonly in the shallower parts of
regional meteoric water flow systems. Many deep-basin brines probably include
re-dissolved halite along with original marine solutes, and perhaps membrane
concentrated constituents as well [14]. A principal example is the Williston-Rocky
Mountain basin of the western USA and Canada, which includes evaporites. The
contribution of solute NaCl to these formation waters has been most systematically
analysed by Hitchon et al. [15]. Brines heavily dominated by NaCl characterize
the Mississippian strata of the Paradox basin [9]; more dilute examples from
California and Texas are given in Ref.[7]. Re-solution of sulphates (calcium
sulphate and polyhalite) can be documented in brines of the Rustler and Castile
Formations associated with the Salado salts of New Mexico [7, 16].
The significance of the compositions of the contained fluids is that the
history of water movements in the salt may be deduced from these data. Sedimen-
tary-basin brines offer evidence of migration from evaporite deposits and of
reaction with other sediments as well. Re-solution brines reflect direct contact
of evaporite deposits with meteoric flow systems. Studies of stable 180 and
deuterium in the water from some major sedimentary-basin brines have
suggested extensive interaction with meteoric fluids and divergent histories for
solutes and solvent [17, 18]. Isotopic analyses of fluids associated with the
Salado evaporites are being examined [8], but complexity and lack of basic data
on many fractionation processes render interpretation equivocal at present.
Similarly, theoretical modelling of saline mineral-brine equilibria for natural
IAEA-SM-243/97 339
fluid compositions is being attempted, but it is severely hampered by the
complexity of interactions in concentrated electrolytes.
2.2. Migration of fluid inclusions in rock salt
Fluids in rock salt tend to be mobilized by thermal energy, by fracture and
decrepitation mechanisms, the dehydration of hydrous minerals, or.the physical
migration of fluid inclusions along thermal gradients. There is much disagreement
about the adequacy of existing data and models [19—22] to predict the amount
and time of arrival of the brine from fluid inclusions at the hot radioactive waste.
The USGS is participating in the design, execution, and interpretation of
Department of Energy sponsored in situ heater tests at Avery Island, Louisiana,
to devise a method for conducting brine-movement experiments using brines
tagged with stable isotopes. The USGS has also served as a consultant for
Department of Energy contractors on brine-movement experiments related to the
Waste Isolation Pilot Plant (WIPP) in New Mexico.
Fluid inclusions from rock salt at WIPP have been described by Roedder et
al. [23]. More than 90 volume per cent of the fluid in this salt is present as
recrystallized primary inclusions larger than 1 mm in diameter (volume
>109 дт3) that show first melting temperatures of — 32° to — 51°C and thus must
contain major amounts of Ca and Mg chlorides. Inclusions larger than ЮОдт
in diameter generally have a small vapour bubble (~0 .1 -0.3 volume per cent)
that homogenizes at 20°-46°C.
The motion of these fluid inclusions in thermal gradients of 0.1° to
2.0°C • mm-1 was measured in the laboratory and found to depend on the size
of the inclusion, the ambient temperature, temperature gradient, vapour/
liquid ratio, and the grain boundaries in the salt. The rate of movement should
also be a function of the composition of the fluid, the presence of non-
condensable gases and organic films, strain, and growth defects and solid
inclusions in the host salt. Detailed laboratory studies [24] have shown that the
three most important factors are ambient temperature, gradient, and inclusion
size. There is no evidence for a threshold temperature for movement of
inclusions.
At 160°C ambient and a gradient of 1.5°C • cm-1, 1-mm diameter inclusions
moved toward the heat source at a rate of 1.7 cm • a-1. Smaller inclusions
moved more slowly; thus, 100 /um inclusions moved only 0.5 cm ■ a-1. The
movement rates at 250°C ambient are at least an order of magnitude larger,
but at 108°C ambient the rates are only 30% lower than at 160°C. In contrast,
natural inclusions with a large vapour/liquid ratio in these same samples moved
at 5 times these rates, but down the gradient, away from the heat source.
These data may be used to calculate the amount of fluid that could be
delivered to the vicinity of the waste canister, and should be compared with
340 STEWART et al.
data from in situ tests. An understanding of the process of movement of fluid
inclusions adequate to explain empirical observations is necessary if we are to
have an adequate basis for predicting the long-term behaviour of fluids in rock
salt.
2.3. Physical chemistry of fluids in rock salt
An understanding of the nature of chemical and physical interactions
between rock salt, its fluids, and the waste canister and waste itself must be
available before a mined repository in salt can be designed for most effective
containment. The basic principles were outlined in Ref.[ 1 ].
The movement of brine in in situ experiments cannot be determined
adequately by collecting and condensing steam and weighing the condensate
[25], because only a fraction of the brine can be converted to steam. Further
more, the steam condensate is nearly pure water, whereas the brine is composed
of water and dissolved non-volatile salts. As the brine boils, it becomes
saturated with solid phases, some of which contain structural water (an example
is tachyhydrite CaMg2Cl6-12H20, which contains 41 weight per cent H20).
The boiling point will be raised as the solution becomes more concentrated, and
eventually no more steam can form at a given temperature and pressure.
The total amount of bittern arriving at the heater can be related to the
amount of steam condensate collected, as follows:
2 b = c - N (1)
1
where 2 b is the total weight of bittern arriving at the heater. N is the weight
of condensate collected, с is a constant for each variety of bittern, / is the
weight fraction of water in the bittern, and x is the weight fraction of the water
that can be converted to steam.
The value of с is unity only for pure water. In the case of a simple salt
where * is approximately unity because of the lack of hydrates and a low boiling
point elevation, the only adjustment to с required is for the amount of salt
dissolved in the brine. For example, for saturated sodium chloride brine at
200°C with a.solubility of 31.898 weight per cent NaCl, с is equal to 1.468. The
value of с is significantly larger for bitterns that can have very high boiling points
and form highly hydrated salts. The value of с might be calculated provided data
were available for the vapour pressures and boiling points of bitterns saturated
with sodium chloride as functions of temperature and bittern composition, the
solubility of NaCl in bitterns as a function of temperature, and the equilibria
IAEA-SM-243/97 341
between bitterns and hydrous salts as a function of temperature and composition.
As these data are generally not available, values of с have been measured directly
as a function of temperature and bittern composition at atmospheric pressure.
In general, the value of с was found to range from 2 to 10 depending on the
temperature and the composition of the bittern. The major controlling com
positional factor was found to be the Mg:Ca ratio.
The above values of с can then be applied to data from brine-migration
experiments to determine how much brine arrived at the heaters. To apply
these data, it is necessary to know the weight ratios of Mg:Ca in the brine, the
approximate total dissolved solids and the temperature at which the bittern is
boiling. The data thus corrected will be significantly greater than many present
estimates of brine that has arrived at heaters. The brine migration equations
currently available will need to be modified accordingly.
Although many of the bulk thermochemical properties of the concentrated
and complex chloride bitterns known to exist in rock salt have not yet been
measured directly, they can be closely approximated from the corresponding
state argument of Potter and Hass [26-28], as discussed with an example in
Ref.fl].
As more components are added to a brine, the greater the mass of fluid will
become at a given temperature for the same amount of H20. Recent USGS
laboratory measurements indicate that SrCl2 and CsCl are both highly soluble in
fluid saturated in NaCl. For example, total dissolved solids in fluid that is
saturated with NaCl and CsCl at 200°C will be 84 grams of salts per 100 grams
of fluid, a greater amount of total dissolved solids than that of combined
saturation in NaCl, KC1, MgCl2, and CaCl2 at the same temperature. The large
effects of depressing liquidus temperatures by adding these chloride components
to the system NaCl—H20 has been discussed in Ref.[l], and both SrCl2 and
CsCl further depress the liquidus.
90Sr and 137Cs contain approximately 70 per cent of the heat-generating
capacity of the fission products in spent fuel aged ~5 years. Only a few litres
of brine in contact with a spent-fuel assembly could dissolve all its contained
strontium and cesium as chlorides. It will be important to ensure that no
possibility exists for such a fluid to form, because it contains its own energy
source and could migrate, thus changing the thermal evolution of the repository
from designed values.
Hydrolysis of bitterns at repository temperatures makes them acidic (to
pH = 2), and they become even more corrosive to metals. We have observed
equilibrium H2 pressures as high as 2 Mpa (20 bars) at 250°C with a NaCl-
saturated Ca, Mg chloride bittern in equilibrium with metallic Ti. Copper metal is
also readily corroded, at rates as high as 40 mm/a at 217°C.
342 STEWART et al.
The preceding sections have indicated that fluids in salt can affect its
performance as a repository medium: thus, it is important to know as completely
as possible the amount and distribution of fluids in evaporites. Measurements
in situ are particularly needed because salt samples readily lose or gain H20
when exposed to the atmosphere.
The USGS has under development new methods using short-wave radar,
hole-to-hole electric resistivity, and borehole proton resonance spectrometry to
ascertain in situ distribution and amounts of H20 in evaporite sections. Vertical
seismic profiling is also being used to determine the structure of beds within salt
anticlines.
New methods under development for dating minerals in salt include a
laser-sampling mass spectrometer for Ar dating, and dating of the Ca produced by
К decay in potassium salts that contain virtually no common Ca.
2.4. New methods to measure fluids in evaporite
REFERENCES
[1] STEWART, D.B., POTTER, R.W., II, “ Application of physical chemistry of fluids in rock salt at elevated temperature and pressure to underlying radioactive waste” , Science Underlying Radioactive Waste Management (MCCARTHY, G.J., Ed.), Plenum Press,New York (1979) 297.
[2] CLYNNE, M.A., POTTER, R.W., И, “ P-T-X relations of anhydrite and brine and their implications for the suitability of anhydrite as a nuclear waste repository medium” , Science Underlying Radioactive Waste Management (MCCARTHY, G.J., Ed.), Plenum Press, New York (1979) 323.
[3] EVANS, H.T., Jr., The thermal expansion of anhydrite to 1000°C, Phys. Chem.Minerals 4 (1979) 77.
[4] DEAN, W.E. (Ed.), Characteristics of Anhydrite, U.S. Geol. Survey, Open-file report (1979), in preparation.
[5] CARPENTER, A.B., Origin and chemical evolution of brines in sedimentary basins, Oklahoma Geol. Survey, Circular 79 (1978) 60.
[6] HOLSER, W.T., “ Chemistry of brine inclusions in Permian salt from Hutchinson, Kansas” , 1st Symposium on Salt, Northern Ohio Geol. Soc., Cleveland (1963) 86.
[7] WHITE, D.E., HEM, J.D., WARING, G.A., “ Chemical composition of subsurface waters” , Data of Geochemistry, U.S. Geol. Survey, Prof. Paper 440-F (1963).
[8] LAMBERT, S.J., “ The geochemistry of Delaware Basin groundwaters” , Geology and Mineral Deposits of Ochoan Rocks in Delaware Basin and Adjacent Areas, New Mexico Bureau of Mines and Mineral Resources, Circular 159 (1978) 33—38.
[9] MAYHEW, E.J., HEYLMUN, E.B., “ Complex salts and brines of the Paradox Basin” ,2nd Symposium on Salt, Northern Ohio Geol. Soc., Cleveland 1 (1965) 221.
[10] POTTER, R.W., II, CLYNNE, M.A., unpublished data (1979), and FERRELL, R., oral communication to B.F. Jones, June 1979.
[11] BRAITSCH, O., Salt Deposits: Their Origin and Composition, Springer-Verlag, New York (1971).
IAEA-SM-243/97 343
[12] SORENSON, H.O., SEGALL, R.T., “ Natural brines of the Detroit River Group, Michigan basin” , 4th Symposium on Salt, Northern Ohio Geol. Soc., Cleveland 1 (1974) 91.
[13] ANDERSON, R.J., GRAF, D.L., JONES, B.F., Calcium and bromide contents of natural waters, Science 153 (1966) 1637.
[14] WHITE, D.E., “ Saline waters in sedimentary rocks” , Fluids in Subsurface Environments, Am. Assoc. Petrol. Geol., Mem. 4 (1965) 342.
[15] HITCHON, B., BILLINGS, G.K., KLOVAN, J.E., Geochemistry and origin of formation waters in western Canada sedimentary basin, III, Factors controlling chemical composition, Geochim. Cosmochim. Acta 35 (1971) 567.
[16] MERCER, J.W., ORR, B.R., Interim Data Report on the Geohydrology of the Proposed Waste Isolation Pilot Plant Site, SE New Mexico, in preparation.
[17] CLAYTON, R.N., FRIEDMAN, I., GRAF, D.L., MAYEDA, Т.К., MEENTS, W.F.,SHIMP, N.F., The origin of saline formation waters, 1, Isotopic composition, Jour. Geophys. Res. 71 (1966) 3869.
[18] HITCHON, B., FRIEDMAN, I., Geochemistry and origin of formation waters in the western Canada sedimentary basin, I. Stable isotopes of hydrogen and oxygen, Geochim. Cosmochim. Acta 33 (1969) 1321-1349.
[19] BRADSHAW, R.L., McCLAIN, W.C. (Eds), Project Salt Vault: A Demonstration of the Disposal of High Activity Solidified Wastes in Underground Salt Mines, Oak Ridge National Laboratory, Rep. ORNL-4555 (1971).
[20] ANTHONY, T.R., CLINE, H.E., “ The thermomigration of biphase vapor-liquid droplets in solids” , Acta Metall. 20 (1972) 247.
[21] GAFFNEY, E.S., Brine Migration, Pacific Technology Rep. PT-U78-0242 (1978).[22] JENKS, G.H., Effects of Temperature, Temperature Gradients, Stress and Irradiation on
Migration of Brine Inclusions within the Salt Adjacent to Waste Packages in a Salt Repository, Oak Ridge National Laboratory Rep. ORNL 5526 (1979).
[23] ROEDDER, E., BELKIN, H.E., “ Application of fluid inclusions in Permian Salado salt, New Mexico, to problems in siting the WIPP nuclear waste repository” , Science Underlying Radioactive Waste Management (MCCARTHY, G.J., Ed.), Plenum Press, New York (1979) 313.
[24] ROEDDER, E., BELKIN, H.E., Migration of fluid inclusions in WIPP salt in thermal gradients, unpublished data (1979).
[25] POTTER, R.W., II., CLYNNE, M.A., THURMOND, V.L., The application of the physicochemical properties of boiling bitterns to the interpretarion of brine migration experiment related to salt repositories, unpublished data (1979).
[26] POTTER, R.W., II, HAAS, J.L., Jr., A model for the calculation of the bulk thermodynamic properties of geothermal fluids, Geothermal Resources Council, Trans. 1 (1977) 243.
[27] POTTER, R.W., II, HAAS, J.L., Jr., Models for calculating density and vapor pressure of geothermal brines, U.S. Geol. Surv. Res. 6 (1978) 247.
[28] HAAS, J.L., Jr., POTTER, R.W., II, The measurement and evaluation of PVTX properties of geothermal brines and the derived thermodynamic properties, Thermophysical Properties (Proc. 7th Symp. Gaithersburg, 1978), Am. Soc. Mech. Eng. (1978) 604.
DISCUSSION
H. GIES: What were the dimensions of the cell in which liquid inclusions
migrate in the direction of the water/source and what were the real contents of
344 STEWART et al.
water in your salt? For example, older rock salt of the Asse mine contains
0.12% water (according to laboratory and in situ studies). However, the main
part of this water comes from the water-bearing accessory salt minerals.
D.B. STEWART: In the experiment illustrated the hot end of the sample
was at 206°C and the cool end was at 198°C about 5.3 cm away. The gradient
thus is 0.15°C/mm. We agree that the dehydration of hydrous minerals will be
additional source of bitterns near the waste canisters.
IAEA-SM-243/13
HYDROGEOLOGICAL RESEARCH AT THE
SITE OF THE ASSE SALT MINE
H. BATSCHE, W. RAUERT
Institut fur Radiohydrometrie,
Neuherberg,
K. KLARR
Institut fur Tieflagerung,
Clausthal-Zellerfeld,
Federal Republic of Germany
Abstract
HYDROGEOLOGICAL RESEARCH AT THE SITE OF THE ASSE SALT MINE.In connection with the storage of radioactive wastes in the abandoned Asse salt mine near
Brunswick (Federal Republic of Germany), the hydrogeology of the ridge of hills of Asse has been investigated. In order to obtain as detailed information as possible on the hydrogeological conditions, a long-term investigation programme has been set up and many methods of investigation have been used. Hydrogeological boring operations resulted in important scientific findings regarding, for example, the course of the salt table and the main anhydrite which towers up above the salt table into the overlying collapsed rocks. Hydrochemical data showed the hydraulic effect of transverse faults. Isotopic hydrological measurements permitted distinction between the flow behaviour of the groundwater in different aquifers. The origin of the salt springs at the northwest end of the structure can be explained. Some additional pumping and labelling tests are expected to yield quantitative results concerning hydraulic interrelationships recognized to date. The very complex hydrogeological structure of the ridge of hills of Asse is the result of the multiple succession of permeable and impermeable layers on the flanks of the structure, and, furthermore, is possibly due to the fact that in some individual faults groundwater may seep through normally impermeable layers as well as via waterways at the salt table.
1. INTRODUCTION
In the abandoned salt mine of Asse the Radiation and Environmental
Research Company, Munich, has since 1967 been carrying out trial storage
operations with radioactive waste material, in order to gain experience in the
technological field of final disposal of radioactive wastes. The pertinent research
work also comprises a ‘Hydrogeological Research Programme Asse’, which has
been set up to investigate conditions within the ridge of hills of Asse.
345
1 1 Aquifer kro Upper ¡> C re taceous m o ( 1,2 ) Upper 'I
•й) Aquitardkru Lower ✓1 m m M iddle ? M uschelka lkmm jb Middle |> Jurassic D o9 9 er ) m u Low er )
1 1 Aquic lude Jl Lower /' (L ias ) so Upper 4|ko Upper 4I sm M iddle \ BuntsandsLeinkm Middle > Keuper P R ogenste inku Low er x1 su
2У
Low er )Zechste in (S a lt-p lu g )
.Cap rock
FIG. 1. Hydrogeological cross-section through the anticline of Asse in the area of Asse II. Simplified representation.
IAEA-SM-243/13 347
P um p ing and lab e lling t e s t s -
Geotopicol field work,examination of drilling samples
Isotope hydrology (Environmental i s o t o p e s ) ___________ _ _
Hydrochem istry of groundwater
Hydrological data of measuring points
¡Environmental monitoring
L is t of m e a s u r in g p o in ts
i— i 1-------1-------1-------1-------1-------- 1----- 1------1-------- r—- 1-------г— г19 6 8 69 70 71 72 73 74 7 5 76 77 78 79 80
FIG.2. The “Hydrogeological Investigation Programme Asse”. Simplified schematic representation. The height of the divisions gives a rough idea of the volume and extent of the work.
2. GEOLOGICAL SURVEY
The Asse is a southeast-northwest striking salt anticline, which lies approxi
mately 15 km southeast of Brunswick in Lower Saxony (Federal Republic
of Germany). On the flanks, the salt uplifting has caused the simultaneous
lifting up of layers ranging from Buntsandstein to the Upper Cretaceous (Fig.l).
At the northwest end of the anticline, the sequence of layers drops under
Quaternary sediments within an area strongly broken by tectonic events, but
with still recognizable centroclinal strike. By virtue of their pétrographie nature,
the flank rocks form an alternating sequence of layers permeable to water and
others more or less impermeable to water (Fig.l). On the southwest flank, this
sequence of layers is dissected by transverse faults [1—7]. In the core of the anti
cline, a salt plug of Zechstein salt appears 200-300 m below ground level (Fig.l).
TABLE I. OUTLINE OF THE HYDROGEOLOGICAL INVESTIGATION BOREHOLES AT THE ASSE
For locations see Fig.3.
B orehole
H
A ltitudea.s.l.(m )
D e p th o fbo ring(m )
F ilte r pipes from - to (m )
D iam eter
(m m )
W ater level 2 Jan . 1979
on g round-surface(m )
a.s.l.
(m )
G eological position o f aqu ife r
1 152.40 70 .0 7 .5 0 - 65 .50 200 - 3.41 148.99 L ow er M uschelkalk
2 143.43 61 .0 1 4 .5 0 - 58.25 125 + 2.32 145.75 K euper and M uschelkalk
3 226 .19 98.0 3 6 .5 0 - 91 .50 150 - 5 6 .6 0 169.69 L ow er B untsandste in
4 2 0 8 .8 9 116.5 1 5 .0 0 -1 1 5 .0 0 125 -5 2 .1 1 156.78 L ow er M uschelkalk
5 188.21 146.0 9 .5 0 -1 1 7 .5 0 150 -1 1 .9 1 176.30 R o t and Z echstein
6 179.98 71 .0 1 5 .7 5 - 68.25 125 - 1 4 .5 8 165.40 Q u a rte m ary and R ô t
7 145.62 50 .0 3 .4 0 - 28 .40 125 + 0 .26 145.88 Lias
8 / 1a 193b 258 .7 2 3 3 .0 0 -2 4 6 .0 0 75 - 6 4 .4 2 128.58 Z echstein
8 /2 193b 180.0 1 4 5 .0 0 -1 8 0 .0 0 50 - 3 9 .4 2 153.58 B un tsandste in (collapsed rocks)
8 /3 193b 120.0 6 0 .0 0 -1 2 0 .0 0 50 - 2 9 .4 8 163.52 B u n tsan d ste in (collapsed rocks)
9 145.1 63.0 2 0 .2 0 - 52 .70 125 - 8 .27 136.83 Lias
10 137.9 38.5 9 .8 0 - 32 .30 125 - 1.41 136.49 U p p er M uschelkalk
11 110.62 149.2 1 2 .2 5 -1 3 5 .2 5 150 + 1.30 111 .92 Q u a ite rn a ry , cap rock , Z echstein
12 114.28 74.0 2 3 .2 5 - 69 .50 125 - 0 .80 113.48 B untsandste in
13 124.54 62 .0 1 4 .5 0 - 59 .50 100 - 1.16 123.38 B un tsandste in
348 BATSCH
E et al.
B orehole
H
A ltitudea.s.l.(m )
D epth ofboring(m )
F ilter p ipes from — to (m )
D iam eter
(m m )
W ater level 2 Jan . 1979
on ground-surface(m )
a.s.l.
(m )
G eological position o f aqu ifer
14 148.6 68.2 8 .8 0 - 4 6 .8 0 150 - 6 .06 142 .54 Q u a rtem a ry , collapsed rocks
15 173.4 50.65 3 9 .9 0 - 49 .90 125 -2 0 .0 5 153.35 M iddle M uschelkalk
15a 175.0 44.1 3 5 .5 0 - 44 .00 125 -2 1 .9 7 153.03 M iddle M uschelkalk
16 178.2 74.0 7 .1 0 - 35 .10 125 - 1 4 .5 0 163.70 M iddle M uschelkalk
17 173.32 243.0 2 2 3 .7 0 -2 4 1 .2 0 125 -4 0 .6 0 132.72 Z ech ste in /sa lt tab le
17a 173.0 242.67 refilled
18 144.5 208 .0 refilled
18a 145.4 230.5 2 0 1 .0 0 -2 1 2 .0 0 150 -2 4 .0 4 121.36 Z echste in /sa lt tab le
19 149.30 214 .50 2 0 4 .0 0 -2 1 0 .2 5 125 -1 7 .2 8 132.02 Z echstem /sa lt tab le
19a 148.97 196.5 5 2 .0 0 -1 6 3 .5 0 125 - 2.35 146.62 B untsandste in (collapsed rocks)
21 148b 190.4 8 .0 0 -1 5 0 .0 0 125 - 8 .00 140 B untsandstem (collapsed rocks)
a H 8 is one bo reh o le w ith th ree observation tubes 8 /1 , 8 /2 , 8 /3 . ^ N o t y e t ex ac tly m easured .
IAE
A-SM
-243/13 3
49
HYDROLOGICAL. EXPOSURES AT THE A S S E
О . Spring, nydrotogicaiiy im p o r ta i
О • S prftg. I *» * important
О • W ell
О • G roundwater gauge(§ ) • Hydrogeological investigation borehote
V a Receiving «ream , вгекчаде, Important
V • Receiving efream. d ra in ag e .ie ** important
* 1 <§>T
сл07 □1 <ÍM lt,neoe*d• . ôeoteg ica l exp loration boring
V2._ , No. of meamurlng po ints
H 1 .H 2 _ . N o.o im veetigatlon bo reho le*
PaPS — a No. oí groundw ater g a ug e*
* Cxpoeure* wUh irrfhjence o f * * t t water
FÏG.3. Layout of the hydrological exposures and of the geological exploration borings at the Asse.
350 BATSCHE
et ai.
IAEA-SM-243/13 351
The horizontal plane at the top of the salt, as a result of subrosion, is termed salt
table (‘Salzspiegel’). (See also Fig.7). Above the salt lie collapsed rocks consisting
of Buntsandstein.
3. HYDROGEOLOGICAL RESEARCH PROGRAMME ASSE
The Hydrogeological Research Programme Asse is diagrammatically
represented in Fig.2. It comprises the following operations:
(1) Survey of the existing springs, wells, drainages and receiving water courses.1
(2) Procuring of the relevant hydrological data as a basis for environmental
monitoring.
(3) Measurement of discharge, water level height, hydrographs, temperature,
electrical conductivity, pH, 0 2-content at the measuring points and at the
investigation boreholes.
(4) Hydrochemical investigation of the groundwaters.
(5) Isotopic hydrological investigation of the groundwaters by determination of
the 3H-, 2H-, 180- and 14C-content, as well as by determination of the MS-
content of the sulphate of selected groundwaters and of sulphate rocks.
(6) Special geological mapping and geophysical reconnaissances. Examination
of the drill cores and cuttings. Construction of weirs for carrying out dis
charge measurements at springs.
(7) Drilling of hydrogeological investigation boreholes in those areas where there
are no or only insufficient hydrological exposures (Table I, Fig.3).
(8) Elaboration of hydrological data in the hydrogeological investigation bore
holes (including single-borehole measurements).
(9) Pumping-, injection- and labelling tests upon completion of the net of
measuring points, with a view to investigating hydraulic interconnections.
In the following, a general summary will be given on the current state of the
exploratory work. Within the scope of this paper, only a few results can be
described in detail. With respect to various results hitherto obtained, conclusive
confirmation will only be possible once the additional pumping and labelling tests
(item 9 of the Hydrogeological Research Programme) have been carried out.
1 These are subsequently called (hydrological) measuring points (1 ,2 ,....), as opposed to the (hydrogeological) investigation boreholes (HI, H 2,....).
352 BATSCHE et al.
4.1. Hydrochemical parameters
According to Frank [7], the Asse area holds autochthonous and allochthonous
groundwaters.
The autochthonous groundwaters move, on their way between their
infiltration area and their reappearance, within a petrographically uniform aquifer.
Their chemical composition is determined only by this aquifer. Allochthonous
groundwaters have flowed through various aquifers or they originate from the
mixing of several groundwaters of different chemical composition. In certain
allochthonous groundwaters, which have been travelling for a long period of time,
a certain percentage of the alkali ions has been exchanged by alkaline earth ions
originating from the aquifer.
The authochthonous groundwaters at the Asse, with the exception of the
Zechstein waters, show chloride contents of less than 150 mg/ltr. The allochthonous
groundwaters at the Asse are always chloridic mineral waters (some 100 mg СГ/ltr).
They are always in direct or indirect connection with fault zones. The chloride
content of the allochthonous groundwaters is derived from the Zechstein salinarium.
On the other hand, it may not be excluded that locally also the salinary facies of
the Middle Muschelkalk and of the Upper Buntsandstein still contain small salt
deposits, thus being able to deliver chloride ions.
About two thirds of all groundwaters at the Asse show mineral contents of
over 1000 mg/ltr; hence they must be termed mineral waters. The mineral
content of the Zechstein waters amounts to more than 200 000 mg/ltr (measuring
point 63, borehole H 11). The fresh water deposits are hydrogen carbonate waters.
As far as the mineral waters are concerned, a distinction may be made between
chloride and sulphate waters. The chloride content originates from the Zechstein,
the sulphate content from the gypsum deposits of the Middle Muschelkalk, of the
Rot and of the Zechstein.
Whenever allochthonous groundwaters with their high-grade chloride content
emerge in the area of the overlying rock beds of the flanks, this may be attributed
to the fact that groundwaters are migrating from the area of the cap rock or of
the salt table to the flank rocks. Owing to the alternating succession of permeable
and impermeable beds surrounding the salt structure (section 2, Fig.l), this
migration, transverse to the strike of the beds, cannot occur except within the
limit of faults causing hydraulic interconnections.
4.2. Iso topic hydrological investigations
4.2.1. The tritium- and carbon-14-content o f the groundwaters
Knowledge concerning the groundwaters in the Asse region has been
supplemented by isotopic hydrological investigations. The 3H and 14C content
of the groundwaters proved to be very important.
4. RESULTS
IAEA-SM-243/13 353
‘Old’ groundwaters which resulted from precipitations which infiltrated
into the subsoil prior to the year 1953, i.e. prior to the beginning of nuclear
weapon tests, show today tritium contents of less than 2 TU2. On the other
hand, ‘young’ groundwater, formed solely from precipitations dating from the
period after 1953/54, today show, on account of the proportion of bomb
tritium in Middle Europe, tritium contents of the order of 100 TU. Whenever
tritium contents of more than 2 TU are found in a groundwater, it may be con
cluded that these waters contain precipitation fractions dating from the period
after 1953/54. Generally, the tritium content is in inverse proportion to the
mean residence time of the groundwater. In the individual case this inter
pretation depends, however, on the model conception underlying the inter
pretation.
The 14C-content of the bicarbonate and the free C02 to be found in the
groundwater originates from the biogenetic soil C02 and, thus, ultimately, from
the atmosphere. According to experience, recent groundwaters show, depending
on the geological parameters of the catchment area, an initial 14C-content of
about 60 to 100% modern3 [8] which might be additionally increased by an
unknown contribution of bomb-produced 14C. The lack of information on the
initial content may be one of the reasons giving rise to an uncertainty on the age
determination. The 13C-content4 may be an indication of the origin of the carbon
in the groundwater and, hence, of deviations from the model conception. (As
for problems involved in 14C-dating of groundwater see for example Refs [9]
and [10].)
For some measuring points, the concentration time curves of the tritium
content are illustrated in Fig.4; with respect to the boreholes, some of the
relative measuring values have been summarized in Table II. The groundwaters
broached by measuring points and boreholes show greatly differing tritium
contents. In the springs, these range between 8 1 TU (measuring point 2) and
24 TU (measuring point 26), whereas in the wells and boreholes they range
between 161 TU (well at Wittmar, measuring point 12) and approximately 0.5 TU
(borehole H 13). Consequently, the proportion of young groundwater greatly
varies from one water specimen to the other. While tritium contents of over or
around 100 TU are characteristic of a groundwater which, for the most part, has
been formed from precipitations occurring during the past 25 years, the measured
2 A tritium unit (TU) is defined as the relation of 3H/*H = 10 ,8. 1 TU = 3.24 pCi/ltr == 0.12 Bq/ltr.
3 The 14C-content is given as modem’ where 100% modem corresponds approximately to the specific 14C-content of living organic substance in the year 1950.
4 Defined as relative %o deviation from a limestone standard (PDB).
FIG.4. Concentration time curves of the tritium content at selected measuring points in the Asse. For location of the measuring points, see Fig.3.2 = spring to the northeast of mine Asse II;20, 21, 26 = springs at the northeast end of the structure;9 = well at Remlingen;12 = well at Wittmar.
354 BATSCHEetal.
TABLE II. TRITIUM CONTENTS (TU, TOGETHER WITH THE TWOFOLD STANDARD DEVIATION) OF THE
GROUNDWATERS APPEARING IN THE HYDROGEOLOGICAL INVESTIGATION BOREHOLES H 1, H 2, H 11 AND H 14
Borehole Date o f sampling
D epth o f sampling
10 m 30 m 60 m 65 m
H 1 14.7.756.6.776.7.77
82 ± 5.6 52.8 ± 3.6
73 ± 5 .1 50.1 ± 3.4 49.7 ± 3.1
72 ± 5 .148.4 ± 3.3
+ 1.3 m 10 m 20 m 30 m 50 m
H 2 1 4 .7 .7 57 .6 .7 76 .7 .7 7
17.7 ± 1.517 .6 ± 1.4
1 6 .6 ± 1.3
17 .8 ± 1.5 17 .8 ± 1 . 8 15 .8 ± 1.8
- 0 m (overflow) 14.5 m 23 m 50 m 60 m 85 m 90 m
H 11 8.2.7218.8.7219.9.73 7 ./1 1.2.75 15.5.75
2.7.7520.6.77
6.7.77
14.2 ± 1.224.2 ± 1.915.9 ± 1.412.9 ± 2.3 17.6 ± 1.5 16.0 ± 1.5 14.4 ± 1.3 14 .4+ 1.5
12.3 ± 2.1
14.1 ± 1.6
13.7 ± 2.2 17.4 ± 1.5. 16.5 ± 1.4
16.5 ± 1.4 16.4 ± 1.3
16.2 ± 1.5
20 m 21 m 47 m 48 m
H 14 9.7.756.6.77 42.8 ± 2.0
53 ± 3.935.5 ± 2.6
17.6 ± 1.5
IAEA-SM-243/13 355
TABLE III. 14C-, 13C- AND 3H-DATA OF GROUNDWATER SAMPLES FROM THE ASSE
Pointof sampling
Date ^-content(TU)
513C(%o)
14C-content (% modern)
Uncorrected 14C-model age3 (years before present)
Measuring point 2 5.7.77 43.2 ± 3.0 -14.7 72.2 ± 1.2 1350 (1200-1500)Measuring point 9< 6.7.77 4.9 ± 0.7 -14.2 57.2 ± 1.0 3300 (3200-3400)
Measuring point 26 6.7.77 2.4 ± 0.3 -12.5 72.7 ± 1.6 1300 (1100-1500)
Borehole H 1 (30 m) 6.7.77 49.7 ± 3.1 -13.3 83.9 + 2.0 100 (0-300)Borehole H 2 (20 m) 6.7.77 16.6 ± 1.3 -12.1 57.7 ± 1.4 3200 (3000-3400)
Borehole H 11 (overflow)
6.7.77 14.4 ± 1.5 -12.9 53.9 ±0.9 3800 (3600-3900)
a Calculated from the 14C-contents, with the assumption of an initial 14C-content of 85% modern. As water samples with a tritium content of more than 2 TU can contain also bomb-produced 14C, the amount of which is not exactly known, the 14C-model ages are greater than those contained in this table, which give only a rough estimate of the age.
356 BATSCHEetal.
IAEA-SM-243/13 357
value of 0.5 ± 0.6 TU corresponds to a groundwater which is free from
atmospheric water from the past 25 years. On the whole, the varying tritium
contents and the 14C-modeI ages (Table III) show a very complex discharge pattern
in the groundwaters of the Asse (see sections 4.3.1 and 4.3.2).
4.2.2. Tritium contents found in the brines o f the mine
It was obvious to investigate whether measurements of the tritium content
of brines in salt mines would give an indication of possible access of young ground
water. As early as 1969, samples obtained from 53 brine points in 15 salt mines
showed, in 62% of all cases, tritium contents of up to 300 TU. The detection
limit of this method was 1.5 TU [11]. In Poland the origin of waters in salt mines
was investigated by measurements of environmental isotopes [12]. Our measure
ments of the tritium contents of the brines in the pits of Asse as well as of the
condensed air humidity of the mine airs, show that the brines in the underground
workings, probably together with the air humidity, absorb tritium from the air to
a large degree. The tritium contents of the brines are clearly higher than those
which used to occur in groundwaters and cannot be attributed to the ingress of
groundwater. (See also Ref.[ 11].)
The correlations between the ventilating system, the tritium contents of the
mine air and of the brines, but also the sampling methods, the influence of the
mine operations on the brines and, finally, the possibility of movements of the
stationary and bore brines in the underground workings cannot be covered in
this short review. On the basis of the investigations made to date, any hydro-
geological interpretation of the tritium contents of brines in salt mines, even in
the case of freshly obtained brines, will not be possible — except in some specific
individual cases - without comprehensive additional investigation.
In the underground workings of Asse II, there has not been any indication
suggesting the presence of a brine which might be influenced by groundwater
originating from the overlying rock beds.
4.3. Hydrogeological aspects of the Asse
Following the first field surveys, the main part of the hydrogeological
investigations was centred on the following four sectors of the Asse: the area of
the large diagonal fault between Remlingen and Gross-Vahlberg (Fig.3); the
area of Wittmar; the area of the northwest end of the structure and, finally, the
area of the collapsed rocks above the salt. The results of the investigation bore
holes showed, additionally, the particular hydraulic relevance of the area of the
salt table.
In the Asse area, average deliveries of 0.5 ltr/s are characteristic of hydro-
logically significant springs, while boreholes having a specific yield of 0.5 ltr/s
t-Л00
M e a s u re m e n t on O c t. 2 5 , 19 73 ; w a te r leve l 0.93 m b e lo w ground s u r fa c e
M e a su re m e n t on Aug.1 5 , 1974 ; w ater leve l 1.01 m below ground su rfa ce Borehole H2
-3 -1E le c t r . c o n d u c t iv it y С 10 S - c m D T e m p e ra tu re С CD
10 11 12 13 1Д 15
10
2 0 -
C ore
30-
¿0C o re -
50
60Depth C m l
P e tro g ra p h y?-Z-~ Clay, silt, calc-tuta
Clay and c laystone,light grey, g re en ish
"£> Si It stone, green, red brown, violet; ^ gyp sum¡Й: Fine grained sa n d sto n e and silt -
___1 stone, red, g re en ish grey, ochrecoloured, redbrow n,brow nish
щ ё ё grey
i j Limestone Trochitenkalk(?), grey ÏT jri C laystone,grey ШТ
Claystone, grey,- gypsum ,black ■ \ \ y f ib ro u s gypsum
Gypsum .blackgrey; c laystone :t V greenish blackgrey, tibrous gyp -
Geologicalposition
Upper?
zс8 5
R a d iu s Cm m l 100 200 300 400 500
- F i l t e r p ipe
- G r a v e l pack
FIG.5. The hydrogeological investigation borehole H 2: Electrical conductivity and temperature o f the groundwater, geological profile andconstruction o f the borehole.
BATSCHE
et al.
M easurem ent on June 25,1974 ¡ water level 6.88 m be low ground surf ace
M easurem ent on Aug- 1 5,19 74 , water level 6.96 m below ground surfaceBorehole H U
-3 -1Electr. conductivity С 10 S -c m 3
1 2 3 4 5 6 7 8 9
Temperature С CJ
9 10 11 12
10 -
C ore -"
20
30-
40-
C o r e -
50-
60
68,2
PetrographyG eo lo g ica lp o s it io n
M a r l,b ro w n is h ye llo w ish
M arlsto ne ,b lue g re y ; sandstone/ed ¡ L im e ston e , ye llo w ish ( m ud )
M arJ .b row n ish g re e n ish ye llo w ish
L im e ston e , b lu e g re y
L im e sto n e W e llen ka lk * g rey M a r l,g re yDo lom ite , ye llo w ish L im e sto n e 'W e lle n ka lk " c la y ish , grey
L im e ston e W e lle n k a lk , grey
Depth Cm]
R a d iu s [m m ]
FIG.6. The hydrogeological investigation borehole H 14: Electrical conductivity and temperature o f the groundwater, geological profile andconstruction o f the borehole.
360 BATSCHE et al.
per metre of drawdown show a relatively rich groundwater occurrence. For the
sake of clearness, Figs 5 and 6 illustrate only a few representative temperature
and conductivity curves. As regards the groundwater of the various measuring
points, investigations carried out to date do not allow the determination of any
correlations between the variations, in time, of the conductivity and the tritium
content.
4.3.1. Hydrogeological aspects o f the overlying rocks
4.3.1.1. The area between Remlingen and Gross-Vahlberg
The most relevant geological element in this area is the zone of the large
diagonal fault (formerly also called ‘large transverse fault’) between Remlingen
and Gross-Vahlberg (Fig.3). Its vertical throw may be set between approximately
300—400 m, the horizontal displacement by 100—200 m, whereby the eastern
block appears to be relatively downfaulted. On the northern flank of the Asse,
towards the western block, the fault zone is sealed by argillaceous and marly
material.
Two springs (measuring points 2 and 3) and two boreholes (HI and H2)
proved the existence of a rich groundwater occurrence in the northeast section
of the fault zone. On account of the tectonic conditions prevailing within the
fault area and the chemical composition of the groundwaters, it may be assumed
that, for the main part, they originate from the eastern block.
Within the northeast area of the fault zone, groundwaters of variable
characteristic features and provenances are being discharged temporally at various
varying rates. Within the area of borehole H 2 (Fig.5), below a level of 42 m, a
medium-strongly mineralized groundwater flows up. At a depth of approximately
40 m, the borehole taps a groundwater with a degree of mineralization which is
about half the previous one, the latter ascending in an artesian mode until above
ground level (Table I). The tritium contents of both waters are equivalent and
practically constant in time (Table II). Within borehole H 1, above a depth level
of approximately 20 m, the mineralization of the groundwater varies over time.
Also, in the course of time the tritium content and the admixture of young
precipitation water vary within the total area of groundwater in this borehole
(Table II). The tritium contents of the groundwaters in both boreholes HI and
H 2 (Table II), related to the 14C-model ages (Table III), indicate a varying dis
charge behaviour. With its tritium contents of 48 to 82 TU and its young 14C-
model age the water found in borehole H 1 shows a large percentage of young
groundwater. On the other hand, the groundwater found in the upper horizon
of borehole H 2, with its 14C-model age of approximately 3200 years, shows,
as far as its main part is concerned, a long mean residence time in a groundwater
reservoir. The relatively small percentage of tritium (about 17 TU) suggests only
IAEA-SM-243/13 361
a small admixture of young groundwaters. Pursuant to these considerations
and the borehole data established (see Fig.5), it may be assumed that the water
tapped in borehole H 1 originates within the area of the large diagonal fault.
On the other hand, the aquifers of the waters emerging in borehole H 2 are
supposed to be located in the layers on the northern flank of the structure, to
the east of the large diagonal fault.
The tritium content of the water found at measuring point 2 decreased, on
average, from 70 TU to 50 TU (Fig.4) between 1969 and 1978. This may be
the result of the decreasing tritium content of the precipitations, as registered
from 1963 to date. Considering the tritium content mentioned above and an
uncorrected 14C-model age of approximately 1350 years, the mean residence
time of the water emerging from measuring point 2 , is between the residence
times of the groundwaters found in boreholes H 1 and H 2. Its mineralization
may probably be derived from the Muschelkalk ([7] p.130). The pertinent
aquifer may be considered to be the Muschelkalk on the northern flank of the
eastern block.
This relatively abundant occurrence of groundwater found in the northeast
region of the large diagonal fault suggests a groundwater flow towards the north
east. On the other hand, in the southwest part of the fault zone, there were no
indications suggesting a groundwater flow towards the southwest. On account
of the results of the investigations carried out in boreholes H 6 and H 16, located
in the zone of the large diagonal fault, there may be possibly only vagrant
groundwaters.
4.3.1.2. The area of Wittmar
This area is of hydrogeological importance, since here the southern flank
of the Asse anticline shows an abundant quantity of faults. It still remains to
be ascertained whether groundwater may penetrate through these faults from
the central part of the structure outwards to the southern flank of the Asse
anticline.
South of the Mine Asse I (Sh 1 in Fig.3), borehole H 14 taps, in the Lower
Muschelkalk, two differently mineralized groundwaters (Fig.6). Approximately
35 m below ground level, groundwater is present, having a conductivity of up
to about 9000 fiS -cm-1. The mineralization of this NaCl-water is derived from
the Zechstein. This groundwater is overlain by a markedly less mineralized
hydrogen carbonate water, with a conductivity of about 1000 ¡iS - cm-1. The
latter groundwater originates from the Muschelkalk area. The tritium contents
of the waters (Table II) suggest certain proportions of young and old ground
water. The sample taken in July 1975 from a depth of 48 m and showing a
tritium content of 17 TU represents the lowest proportion of young water.
362 BATSCHE et al.
The former drinking water well of Wittmar (measuring point 12), located
in the Rat sandstone, takes up a slight inflow quantity of chloridic Zechstein
water. Hence, it carries allochthonous groundwater [7]. At the same time, the
water shows a particularly high tritium content (Fig.4). Consequently, two
possibilities of hydrogeological interpretation may be derived:
(a) Either the well, located in the Rat sandstone area, receives precipitation
water seeping in the central part of the structure and taking up a certain
amount of salt prior to arriving at the well after a relatively short flow time;
(b) or the well is being fed, essentially, by young precipitation water, seeping
in the Rat sandstone area of the southwest flank, and to which is being added
a small amount of a somewhat stronger concentrated salt solution, originating
from the interior structure area, after a relatively long flow time.
The chemical composition of this specific groundwater suggests a relatively
long migration path of the NaCl water, the Na being substituted for a Ca, the second
interpretation thus being more likely to hold true.
Within the area of Wittmar the groundwaters made accessible through wells
and boreholes show a certain influence of chloridic Zechstein waters originating
from the collapsed rocks or the cap rock and arriving at the flanks of the Asse
anticline. In the Lower Muschelkalk, i.e. in the inner part of the structure, this
influence is evident; it is already very slight in the Rat sandstone of the Upper
Keuper (see also Fig. 1 ).
4.3.1.3. The northeast flank of the structure
On the northeast flank of the Asse, between Gross-Vahlberg and the north
west end of the structure, the overlying rocks do not contain any faults. There
are, moreover, hardly any emergences of groundwater. Only in the vicinity of
Mônchevahlberg, does a medium-sized spring emerge and to the west of Gross-
Vahlberg, a moderate spring (measuring points 40 to 29). At measuring point 41,
there is only a temporary small discharge of drainage and gravitational water
(Fig.3). On the northeast flank of the structure, there are no indications suggesting
an infiltration of groundwater from the interior of.the structure to the flank area.
According to the results of the investigations, the northeast flank of the Asse is
hydraulically impermeable with respect to the central part of the structure.
4.3.1.4. The northwest part of the structure
With only a few exceptions, the springs emerging at the Asse are located
at the northwest extremity of the structure (Fig.3). Almost the entire overground
portion of the groundwater discharge emerges here. The groundwater flow thus
occurs predominantly along the strike of the various aquifers.
IAEA-SM-243/13 363
Some of the springs on the northwest extremity of the structure are highly •
mineralized salt springs with mineral contents of more than 200 g/ltr (measuring
points 52, 63). Other springs reveal a more or less intensive salt water influence
(Fig.3). Areas showing rudimentary plant growth on the fields between Gross-
Denkte and the northwest extremity of the structure, as well as seepage water
originating from the drainages indicate salted groundwater (section 4.3.2).
At the northwest extremity of the structure, the oolithic “Rogenstein” of
the Lower Buntsandstein and the Limestone of the Lower Muschelkalk, in the
overlying rocks at the northern flank, constitute groundwater aquifers. From
both horizons, important springs come forth in this area (measuring point 26 in
the oolith bed, delivery 0.3—0.9 ltr/s and measuring point 20 in the Lower
Muschelkalk, delivery 0.3-2 ltr/s). In the area of mine Asse II, the groundwater
flow in both horizons is still only scanty, as was found by boreholes H 3 and H 4
which, at this end, have been sunk in the oolith and the Lower Muschelkalk
respectively. In borehole H 4, which penetrates the Lower Muschelkalk approxi
mately 100 m, a preliminary test carried out in connection with the sinking of
the borehole permitted, on a drawdown of 75 m, an extraction of only 0.2 ltr/s.
Thus, some horizons carry no groundwater in the southeast part of the structure,
whereas they carry a large volume of groundwater at the northwest end of the
Asse anticline. At least on the northeast flank of the Asse anticline, conversion
into an aquifer takes place, in a few horizons, within the strike of the beds from
southeast to northwest.
The water of the oolith spring (measuring point 26) shows a 14C-model age
of approximately 1300 years (Table III). The tritium content amounts, with only
one exception, to 3—16 TU, thus being very low (Fig.4). Consequently, the
groundwater in the horizon of the oolith is being discharged via a reservoir, where
the main part of the water is being exchanged only very slowly. This old ground
water portion receives, as a rule, only a slight admixture of young groundwater,
the amount varying over time.
The Muschelkalk water discharging at measuring point 20, shows tritium
contents between 25 and 60 TU, receiving an also varying but decidedly larger
admixture of young groundwater, as compared to the water of the oolith spring.
Most of the groundwater is discharged here relatively rapidly. As a consequence
of this, the Muschelkalk shows a higher water conductivity than the oolith.
4.3.1.5. The collapsed rocks
The collapsed rocks lying over the salt core of the structure consist of
blocks of Lower, Middle and Upper Buntsandstein. These blocks tipped more
or less towards each other. A predominant part is played here by barely permeable
and impermeable rocks. On fractures and joints the rocks carry groundwater.
ыON4
L e ge n d :S. * W a te rle ve l in t h e b o re h o le s — * P ie z o m e t n ç (evet of t h e b r in e s a t the so it tab le Q = Q u a te rn a ry/ , » C o lla p se d r o c X s and cap ro c k
(s a n d s to n e , c lay , g y p s u m )
> ‘ G y p su m ; c.c = c o m p a c t g y p su m la ye r of cap ro c k л * A n h y d r ite ( A 3 “ Main anhydrite)J = SoitMil = C lay (T 3 = Grey soit clay)/ ' Fault
/ = Fault, s u p p o s e d
• : F ilfe r pipeS h = Sh a f t (w ith N o )H = H y d ro g e o lo g ic a l in ve s tig a t io n b o r e h o le (w ith N o )R = G e o lo g ic a l in v e s t ig a t io n b o re h o le (w ith No.)
«>HСЛПXИ
H B a t s c h e К K la r r 1 9 7 9
FIG. 7. Longitudinal section through the Asse in the area of the collapsed rocks above the salt. For location of the boreholes see Fig.3.
IAEA-SM-243/13 365
The inhomogeneous structure, in connection with blocks which are hydrauli-
cally sealed with respect to each other, obstructs the movement of groundwater.
In the area of collapsed rocks, there is no continuous groundwater body. The
water levels which build up in the boreholes are influenced by the inhomogeneous
structure of the rocks. Thus, a difference of about 10 m is shown between the
water levels in observation pipes 8/2 and 8/3, which both belong to borehole H 8 (Fig.7, Table I). Therefore, any groundwater level gradient between different
boreholes in the area of the collapsed rocks cannot be determined except by
reference to the specific geological structure.
4.3.2. Hydrogeological aspects in the area o f the salt table
Figure 7 illustrates a longitudinal section through the Asse in the area of
the collapsed rocks, as shown on the basis of the results of the hydrogeological
investigation boreholes. The specific area which, in the section illustration, has
been marked as collapsed rocks, contains inclusions of gypsum which, as residual
deposits, may be attributed to the cap rock. But since also the salinary facies
of the Rôt (Upper Buntsandstein), which may be found in the collapsed rocks,
contains gypsum deposits, there has been no possibility to date safely to delimit
the cap rock from the hanging layers. On the other hand, the lower part of the
cap rock reveals itself clearly as a compact gypsum horizon of a thickness of
12—17 m, which directly overlies the salt plug.
In general, the salt table runs almost entirely in a horizontal direction, being
located at about 55—60 m below sea level. Within the area of the shaft Asse 2,
it dips about 40 m (Fig.7). At various points, the main anhydrite of the Zechstein
rises in different amounts above the salt table, towering up into the collapsed
rocks (Fig.7, boreholes H 5, H 17+ H 17 a, H 8 ,H 11).
O f the six boreholes which were drilled down to the top o f the salt plug(H 8 , H 17, H 17a, H 18, H 18a, H 19), three (H 17, H 18, H 19) tapped,
directly at the salt table, i.e. between the salt and the compact gypsum horizon,
brine-carrying flumes [13]. Television investigations showed that the vertical
opening width of these flumes is about 5—10 cm, whereas no knowledge has
so far been obtained as regards the horizontal extension. Between the flumes,
the salt table is dry. Here, the gypsum horizon represents a compact bed lying
on the salt surface (boreholes H 17 a and H 18a; in the case of borehole H 18a
a systematic pressing-in of fresh water established a connection to the flumes at
the salt table and this borehole is today used as a observation well.) Within
borehole H 8 no flumes were found underneath the cap rock.
The piezometric level of the brine in the flumes is found approximately 180
to 190 m above the salt table and shows a decline from southeast to northwest
(Fig.7, Table I). Between boreholes H 17 and H 19 the difference in the water
levels is only about 70 cm, corresponding to a gradient of 1 %o . Between
366 BATSCHE et al.
borehole H 19 and borehole H 11 the gradient is approximately 9%o . On account
of the hydraulic pressure conditions it may be supposed that groundwater will
seep down to the salt table in the area around borehole H 17. The main anhydrite
portion which, in the area of borehole H 17, extends up to ground level, is strongly
karstified and may thus be effective as an inflow path of groundwater. This
groundwater of the brine originating from it flows northwest from there, via the
flumes at the salt table. To the east of Gross-Denkte, the piezometric level
crosses the ground level. Here, the brine rises, forming salt springs and salted
groundwater. Borehole H 11 has tapped this ascending brine within the karstified
main anhydrite (Fig.7). The piezometric level here lies more than 1 m above
ground level (Table I). By means of an injection test carried out within borehole
H 17, a transmission of pressure could be observed to borehole H 11, affecting
also boreholes H 19 and H 18a. This connection existed via the brine-carrying
flumes at the salt table. Altogether, the results of these investigations explain
the origin of the salt springs near Gross-Denkte, whose existence has long been
known. The flow movement, as derived solely from the drop of the piezometric
level, has still to be proved by labelling tests (item 9 of the Hydrogeological
Research Programme).
The brine found in borehole H 11 shows a 14C-model age of approximately
3800 years (Table III). This indicates that the hydraulic phenomena occurring
within the collapsed rocks and at the salt table are being determined by a prolonged
medium residence time of the groundwater in the reservoir of the fissures and
eventually the pore space. A tritium content of 12—17 [24] TU (Table II) shows
that the brine contains small amounts of precipitation water dating from the past
25 years. Within borehole H 11, the tritium contents are equal across the
vertical profile and particularly below and above the inflow at about 16 m. The
inflow of young groundwater consequently does not occur within the borehole,
but within the catchment area of the groundwater forming the brine.
In the geological exploration boring Remlingen 4 (R 4) and in the shaft
Asse 2, the compact gypsum horizon on the salt table could no longer be traced.
As the origin of this gypsum horizon depends on the brine-loaded flumes at the
salt table [13], it must be assumed that in the area around the shaft Asse 2, there
will be no flumes at the salt table with connection to the western part of the Asse.
In the boring R 4 the cap rock and the salt table were found to be completely
dry. The culmination of the piezometric level of the brines may be expected in
the area around borehole H 17, which is located 1 km from the shaft Asse 2.
The gypsum horizon opened up between boreholes H 17 and H 18a consists,
according to the results of the borings, of compact gypsum and is not karstified.
In order to be quite sure about these phenomena it will however be necessary to
undertake prolonged observation of the fluctuations of the groundwater level in
the collapsed rocks area.
IAEA-SM-243/13 367
The hydrogeological characteristics of the Asse are determined by the
alternation of permeable and impermeable beds within the overlying rock of the
flanks, by faults crossing this series of strata in several areas and finally by water
ways at the salt table.
The alternating succession of permeable and impermeable strata provokes
the groundwater movement in the aquifers to follow, essentially, the strike of
the beds, from southeast to northwest until the northwest extremity of the
structure. To the east of the shaft Asse 2, in the area of Wittmar and at the north
west extremity of the structure, there are hydraulically effective faults, running
transversely to the strike of the beds. Locally, such faults may entail a slight water
movement even in the aquitards and aquicludes.
The area of the large diagonal fault located to the east of the shaft Asse 2
shows to the northeast a relatively rich groundwater occurrence. Here ground
waters are being discharged which are different in their chemical composition
and in their relative 3H- and 14C-contents. They are partly generated in the fault
area itself and originate partly from the eastern block. In the direction of the
western block, i.e. in the area of the shaft Asse 2, the fault zone is, according to
the existing investigation results, hydraulically impermeable.
In the area of Wittmar, the southern flank of the structure is tectonically
faulted. The chemistry of the groundwaters in this area shows that within the
faults a certain amount of groundwater may seep, from the central part of the
structure, into the overlying rocks, but does not reach beyond the flanks of the
structure.
Along all the northeast flank, from Gross-Vahlberg to the northwest end
of the structure, there are no faults whatsoever. Neither are there any indications
suggesting a groundwater movement transverse to the strike of the rocks. The
northeast flank may be regarded as hydraulically impermeable.
Almost the entire overground groundwater discharge from the Asse proper
(i.e. to the west of the large diagonal fault) occurs at the tectonically faulted
northwest end of the structure.
In the area of the salt table, a certain amount of brine flows in flat flumes
below a probably sealing gypsum horizon, from the area around borehole H 17,
to the northwest extremity of the structure. Here, it rises to the surface, producing
the salted groundwaters to the east of Gross-Denkte. In this way the origin of the
salt springs at the northwest end of the structure can be explained. On the other
hand, in the area around shaft Asse 2, according to the results of the investigations
the salt table is dry.
The aquicludes and aquitards at the flanks of the structure, which, in some
specific cases, have a thickness of up to 200 m, represent groundwater barriers.
At a few individual points, along faults, there is probably a seeping movement
5. SUMMARY AND CONCLUSIONS
368 BATSCHE et al.
of groundwater through the aquicludes and the aquitards. Within the area of such
faults, the intercepting effect of the impermeable and hardly permeable layers
appears to be reduced. The hydraulic effect of these faults does not extend beyond
the flanks of the structure. On the whole, aquicludes and aquitards form, for most
of the Asse, a divide between the groundwater in the interior of the structure
and the environmental area.
Due to the existence of different aquifers and to hydraulically effective
faults, the hydrogeological phenomena of the Asse display a great variety of
aspects. A detailed knowledge of the hydrogeological conditions, therefore,
required a long-term programme of investigation and the use of various research
techniques. A quantitative evaluation of the hydraulical interconnections found
to date in the course of the investigations will thus be possible only when the
results of the pumping and labelling tests become available.
But even if the hydrogeological investigations alone do not allow conclusions
concerning the safety of the radioactive waste deposits in the Asse salt mine, they
nevertheless represent an indispensable basis for general safety studies carried out
in connection with the storage.
ACKNOWLEDGEMENTS
We wish to express our appreciation to Mr. Albrecht, director of the Technical
Department of the Institut für Tieflagerung, and Mr. Kolditz and Mr. Opp for
managing the financial part of the bore programme. We also thank Mr. Thielemann,
manager of the Asse mine, for the valuable co-operation extended in connection
with various field operations.
We would specially like to thank all members of the Institut für Tieflagerung
and the 3H- and 14C-laboratory of the Institut für Radiohydrometrie who have
carried out numerous measurements in the field and in the laboratories.
The costs of the hydrogeological bore programme were met by the Bundes-
minister für Forschung und Technologie.
REFERENCES
[1 ] HARBORT, E., Geologische Karte von Preufien 1:25 000, Bl. Wolfenbiittel 2094 (Gradabteilungs-Blatt 3829) Berlin (1931).
[2] WOLDSTEDT, P., Erlâuterungen zur Geologischen Karte von Preufien und benachbarten deutschen Lândern, Bl. Wolfenbiittel 2094; mit einem Beitrag von G. Gorz, Berlin (1931)
[3] WOLDSTEDT, P., HARBORT, E., Geologische Karte von Preufien 1:25 000,Bl. Schôppenstedt 2095 (Gradabteilungs-Blatt 3830) Berlin (1931).
[4] WOLDSTEDT, P., Erlâuterungen zur Geologischen Karte von Preufien und benachbarten deutschen Lândern, Bl. Schôppenstedt 2095 mit Beitrâgen von E. Fulda und G. Gorz,Berlin (1931).
IAEA-SM-243/13 369
[5] KALKA, H., Tektonische Analyse des Asse-Heeseberg-Zuges. Dissertation TH Braunschweig, Brunswick ( 1963).
[6] APPEL, D., Bericht über die geologische Neuaufnahme der Asse bei Wolfenbuttel (Ost- niedersachsen) unter besonderer Berücksichtigung ihrer Quartârbedeckung und der Tektonik der Siidwest-Flanke, Diplomarbeit, TH Braunschweig, Brunswick (1971).
[7] FRANK, H., Hydrogeologische und Hydrogeochemische Untersuchungen an der Asse bei Wolfbiittel, Dissertation Universitàt München 1972, Gesellschaft für Strahlen- und Umweltforschung Bericht R 87, Neuherberg (1974).
[8] GEYH, M.A., “On the initial 14C content in groundwater” , Radiocarbon Dating, (Proc.8th Int. Conf. Wellington, 1972) Paper D 58-D 69.
[9] MÜNNICH, K.O., Isotopen-Datierung von Grundwasser, Naturwissenschaften 55 (1968) 158-163.
[10] Isotope Hydrology 1978, IAEA, Vienna (1979).[11] GEYH, M.A., “Messungen der Tritium-Konzentration in Salzlaugen”, Kali und Steinsalz
5(1969) 208.[12] ZUBER, A., GRABCZAK, J., KOLONKO, М., “Environmental and artificial tracers for
investigating leakages into salt mines”, Isotope Hydrology 1978, Vol.I, IAEA, Vienna (1979) 45-63.
[13] BATSCHE, H., KLARR, K., Gedanken und Beobachtungen zur Gipshutgenese, (Rep.5th Int. Symp. on Salt, Hamburg, 1978) (in press).
DISCUSSION
P. FRITZ: Is it possible to use your hydrogeologic and geochemical data
to estimate the rates of salt removal from the salt dome? If so, have you
attempted to do any quantitative calculations?
K. KLARR: Because of the local distribution of the measuring points, a
quantitative calculation of the rates of salt removal is not yet possible at places
where high groundwater salinities have been determined. However, the measure
ments show that the leaching of salt occurs only in some limited sections along
the Asse mine area under investigation.
Valentina BALUKOVA: Do you have data on the stability over time of
mines in salt formations?
K. KLARR: There is a large body of data in the literature. However, the
purpose of our paper was to describe the present hydrogeological situation in
the environment of the Asse salt mine.
G.V. EVANS: The values of tritium measurements in Table III show only
qualitatively the levels of modem waters mixed with older groundwaters at your
sampling sites. In such situations considerable uncertainty must be associated
with any interpretation of 14C measurements for groundwater dating. In addition,
there are other uncertainties due to the supposition of the initial 14C values and
geochemical corrections. In view of these points, could you please explain the
370 BATSCHE et al.
derivation of the narrow tolerances given for 14C ages quoted in Table III? Do
you think it wise to base hydrological interpretations on differences between
these age values?
K. KLARR: It is true that the tritium contents provide information on the
admixture of young waters. This and also other uncertainties of 14C age inter
pretation give rise to difficulties in determining 14C age. The model ages given
in Table III have to be taken as minimum values for the long-term component
of the groundwater investigated. Our hydrogeological interpretation takes these
points into account.
IAEA-SM-243/16
DISPOSAL AND FIXATION OF MEDIUM- AND
LOW-LEVEL LIQUID WASTES IN SALT CAVERNS
In situ solidification
R. KÔSTER, R. KRAEMER, R. KROEBEL
Nuclear Research Centre,
Karlsruhe,
Federal Republic of Germany
Abstract
DISPOSAL AND FIXATION OF MEDIUM- AND LOW-LEVEL LIQUID WASTES IN SALT CAVERNS: IN SITU SOLIDIFICATION.
The paper describes a new concept for the treatment and final disposal of liquid medium- and low-level wastes as produced at the projected FRG nuclear fuel cycle back-end centre, which comprises waste-producing facilities and waste disposal in a salt formation on the same site. The main features of the concept are: (1) préfabrication of granules (diameter about 10 mm) with the aqueous medium- and low-level wastes and inorganic binders above ground;(2) transportation of the granules in cementitious grout via a vertical pipeline into a 75 000 m3 cavern located about 1000 m below ground. The grout may use tritiated water; (3) in situ solidification of the concrete-like waste product in layers. This concept offers a number of advantages as regards technical, safety and economic factors. The necessary data for establishing the technical concept and the safety analysis will be elaborated by mid-1981. The R & D work is sponsored by the Federal Ministry for Research and Technology and is being executed by a number of institutions. The main R & D areas are the following: (1) specification of raw wastes; (2) préfabrication of pellets and determination of their properties: e.g. abrasion and point pressure resistance; (3) behaviour of the reference product (pellets mixed with cement grout) during transport and the filling stage; (4) character of the final product with respect to the leach resistance of mainly Cs, Sr and tritiated water, mechanical properties, corrosion resistance, production of radiolysis gases; (5) the cavern system’s construction and geomechanical stability; (6) technical equipment, i.e. pellet fabrication and mixing of cement grout, transportation concept via vertical pipelines and off-gas system; (7) safety analysis concerning: temperature rise in the product due to hydration; heat release; contamination of the cavern atmosphere.
1. INTRODUCTION
This paper describes a new concept for the treatment and final disposal of
low- and medium-level liquid wastes arising as generated at the projected Federal
German nuclear fuel cycle back-end centre (NEZ).
The NEZ comprises the following nuclear facilities: fuel element interim
storage; reprocessing plant (1400 t U/a); uranium conversion plant; mixed oxide
371
372 KÓSTER et al.
FIG.L Reference system for the container-free disposal of liquid MLW/LLW in a salt cavern.
fuel element production facility; waste treatment and final repository in a salt
formation deep underground at the site.
The basic idea for the new concept results from the concentration of waste-
producing facilities and the final repository at one site, as there is no need for
waste transport on public roads or railways using shielded waste containers.
The consequence is the idea of container-free disposal using fluid product
during transportation which is hardened in situ in the final disposal area.
2. DESCRIPTION OF THE CONCEPT
The wastes arising as described in Section 3 are granulated above ground by
adding inorganic binder with the help of special technical equipment as described
in Section 4 (see Fig. 1 ). The prefabricated pellets are mixed with a cement grout
using either inactive water or tritium-enriched water as generated at the high-level
waste (HLW) concentration and vitrification facility. This mixture is transported
IAEA-SM-243/16 373
FIG.2. Scheme of three caverns in different states: construction, operation, decommissioning.
by gravity through a vertical pipe directly into a big cavern with a capacity of
75 000 m3 located 800—1000 m underground. The velocity of the mixture in
the vertical pipe is controlled by the wall friction and the pipe diameter.
The pipe outlet is located at the bottom of the cavern where the mixture
spreads out, thus forming a layer which hardens in situ. The free pipe end in the
cavern is shortened as the level of filling rises.
Each layer represents one batch and it is expected that nearly 100% of the
available cavem-space can be utilized. One cavern can hold the liquid wastes
arising from 5 years’ operation of the NEZ.
Figure 2 shows a scheme representing three cavern systems in different
states: construction, operation and decommissioning. Each cavern is made after
drilling a borehole either by solution mining or by dry mining techniques. The
caverns are vented during operation in order to maintain defined pressure condi
tions. After the filling operation the remaining borehole at the top end of the
cavern is sealed.
374 KÔSTER et al.
TABLE I. CHARACTERISTICS OF SOLIDIFIED LIQUID MLW AND LLW
WITH CEMENT
Waste type Waste category Final product
(m3/a)
Mean|3/7-activity(Ci/m3)b
1.1 Distillates 5 500 130 ■(used as filling containing (water/cementmaterial) НТО ratio = 0.4)
1.2(used for pellet
Aqueous low- level waste
2 700a 8 X 10~2
production) Aqueous medium- level wáste
5 300a 260
Ashes 140 1
Bead resins 750 25Powder resins plus evaporator concentrates
820 80
a 10% salt content.b Average a-activity = 1СГ3 Ci/m3 for waste type 1.2.
Compared to conventional techniques for the disposal of waste drums in
man-operated underground salt chambers this non-man-operated disposal concept,
as described above, offers a number of advantages as regards safety and economic
and technical factors. For instance:
large quantities of wastes can be handled easily;
the small number of processing steps, lack of container, and simplified
handling will lead to a reduction of radiation doses to the staff, which works
only above ground;
this concept should achieve complete filling of the caverns, resulting in a
final state of the salt formation which is very close to its original state;
the formation of a quasi-monolithic block with a small surface-to-volume ratio
is of advantage in case of a brine intrusion;
IAEA-SM-243/16 375
the utilization of the cavern volume is about ten times better than if containers
with lost-concrete shielding are used; thus relatively low costs may be
achieved. Preliminary cost estimates have shown that the proposed concept
decreases specific costs by a factor of 5 compared with the conventional
concept.
In order to solve R & D problems arising from this concept the following
items are being investigated :
specification of the wastes arising in the NEZ; quantities, chemical and
radiochemical composition;
determination of waste-binder properties in the transport/filling stage:
viscosities, setting time of cementitious suspensions;
determination of pellet properties: abrasion, point pressure resistance;
characterization of the final product in the salt cavern: leachability of
mainly Cs, Sr, НТО; mechanical properties, corrosion resistance, radiation
resistance;
the cavern-system itself: selection of conditioning, transport criteria, analysis
of the stability of caverns, calculations of temperature rise due to hydration
heat and decay heat, radiolytic gas production, safety assessments;
studies of the basic engineering features: dosing and mixing, pelletization,
transport through vertical pipes, off-gas treatment.
The project started at the end of 1977 and will last till July 1981. During
this time the necessary data will be elaborated for establishing the technical
concept and the safety analysis.
3. WASTE SPECIFICATION
3.1. Primary wastes
The nuclear fuel cycle back-end centre (1400 t U/a) will produce a variety
of concentrates and distillates. Their final volumes after solidification by
cementation and mean specific /З/7-activities are listed in Table I.
Detailed characteristics of various raw waste categories, such as radionuclide
concentration and chemical composition, based on data from WAK (Wieder-
aufarbeitungsanlage Karlsruhe), have been elaborated. About 30% of the total
activity comes from 137Cs, 24% from 106Ru/Rh, and 10% from 90Sr in 5-year-old
medium-level concentrates from the reference fuel at a bum-up of 30 GW-d/t.
The mean specific alpha activity of waste type 1.2 is about 10-3 curies per
cubic metre of final product.
376 KÓSTER et al.
The required volume of the underground cavities is dependent on several
conditions:
(1) net volume of wastes arising as specified in Table I;
(2) the reference product is represented by waste type 1.2 in the form of granules
or pellets (prepared above ground) which are fixed in the cavern by cement
grout prepared with waste type 1.1.
Waste type 1.2 requires a net volume of about 10 000 cubic metres per year
(10% by weight of salts in the pellets). The free space between the pellets yields
roughly a volume of about 5000 m3/a. This corresponds to the amount of tritium
waste water fixed in cement (water-to-cement ratio 0.4:1). The total space
requirement thus amounts to about 15 000 m3/a.
4. PELLET PRODUCTION
The separation of fixing into two steps, i.e. conditioning (pellet production)
and filling-up the free spaces, offers a number of advantages:
(1) The binder composition can be fully adjusted to the waste characteristics
at the conditioning step (pellet formation). This means in particular that the
materials used, such as bentonite for cesium retention and densifying
additives for strontium retention, can be applied effectively.
(2) The conditioning above-ground allows techniques suitable for high-density
products to be used, this being an essential precondition for long-term corrosion resistance.
(3) The material for filling the spaces can be much more easily adjusted to the
requirements, i.e. good flow characteristics, good long-term corrosion
resistance, low heat of hydration and slow generation of heat of hydration.
(4) Compared to the introduction as a suspension, pellets release notably lower
heat to the surrounding formation, because the heat of hydration is already
generated during the processing step above ground.
(5) Quality control on pellets above ground can be performed much more
effectively.
The concept for conditioning by pelletization is relatively new in this project.
Therefore only preliminary considerations concerning its technological realization
are available. Pellet production using cement and simulated medium-level waste
solution has been carried out so far on a laboratory-scale facility with a batchwise
operating rotating dish.
In the runs performed so far the influence of various parameters on the
pellet formation has been investigated, such as speed of rotation, as well as the
inclination of the dish. The dish inclination has a great effect on the pellet size.
IAEA-SM-243/16 377
water/cement mixturerproportion of water-----►
FIG.3. Pellet size compared with water /cement ratio.
Monodispersed pellets of about 10 mm diameter are formed at an angle of 45°
to 47°.
In Fig.3 the pellet size is plotted against the water/cement ratio. Optimal
pellets are found in region c. In region a, the liquid content is not sufficient,
whereas in region d, it is too large, thus the pellets disintegrate due to their own
weight.
The abrasion resistance of various pellet products was compared under
standardized conditions. Pellets containing bentonite for cesium retention show
a considerable increase in abrasion resistance.
The pellets were also subjected to fall and fracture tests. The fracture strength
increases with the age of the pellets, as shown in Fig.4. The experiments are
designed to shift the increase of the strength towards shorter times by using
special types of cement.
A planning study for a hot prototype laboratory pelletization facility using
an inclined rotating dish has been carried out. The technical flowsheet inclusive
layout of instrumentation is based on the following data:
378 KOSTER et al.
Age of pellets (h)
FIG.4. Point pressure resistance plotted against age of pellets.
Material data
Cement :
Particle size
Apparent density
Waste concentrate:
Solution containing
Salt content
Density
Pellets:
Diameter
Weight
Cement content
Concentrate
10—100 jum
1.5 kg/ltr
20% NaN03 220 g/ltr
1.13 kg/ltr
10—12 mm
1.2—2.1 g
81 wt%
19 wt%
IAEA-SM-243/16 379
Throughput data
Throughput cement
Throughput concentrate
Dish inventory
Residence time in dish
Rolling speed of pellets
Throughput pellets
750 kg/h
170 kg/h
47 ltr
6 m in3.8 m/s
920 kg/h
5. IN SITU SOLIDIFICATION PROPERTIES OF THE FINAL
MONOLITHIC PRODUCT
The main efforts on the optimization of the reference product so far were
spent on homogeneous solidified MLW/LLW and cemented tritiated water
respectively, whereas the properties of the quasi-monolithic block formed from
the granules and the filling material have not been determined experimentally as
yet. This block can be considered as a concrete-like product, the granules being
a substitute for the gravel. The strength of the granules is relatively small but
increases with time (see Fig.4). The strength of the filler, consisting only of
inactive components such as cement, sand and possibly clay, will certainly be of
the same order of magnitude or higher.
The optimization criteria for the reference product were the following:
reduction of leachability of T, Cs, Sr
long term stability
mechanical strength
flow characteristics
pelletization behaviour.
As already reported in Refs [1, 2,3] the leach resistance of Cs and Sr could be
improved by a factor of ~ 60 and ~ 10 using bentonite and inorganic cement
additives respectively.
6 . TECHNICAL INSTALLATIONS
6.1. Vertical transport of the waste into the cavern
The selected concept for the vertical transport of the pellets mixed with grout
cement is characterized by the following items (see Fig.l):
(1) the concrete-like product flows by gravity through the vertical pipe from the
filling station above ground into the cavern at 800—1000 m underground;
380 KOSTER et al.
(2) the velocity of the fluid product is controlled by the geodetic pressure of the
liquid column and the wall friction, which are in equibalance:
pB -g-H = 4 H/D (a + v-b)
Pg = mean density of the mixture (kg-m-3), g = 9.81 m-s~2,
H = pipe length (m), D = pipe diameter (m),
a = inertial movement resistance (N • m-2),
b = velocity factor (N • s • m-3),
v = velocity (m-s-1).
The viscosity of the concrete-like mixture is an important factor for the
resulting velocity. It can be adjusted by certain additives in the cement grout.
With preliminary data taken from the literature the main characteristics of
this transportation concept are as follows:
pipe diameter 150 mm
velocity 0.1 m/s
throughput 6.4 m3/h
operating time per year 2400 h
(3) The relatively low velocity of the mixture largely prevents any erosion of
the wall, which is the most critical factor in this kind of transportation system.
(4) The presence of pellets in the mixture eliminates the necessity for special
cleaning procedures of the pipe as any potential generation of layers at the
inner pipe wall will be removed by the pellets themselves. The throughput
is permanently controlled in order to detect any kind of malfunction of this
system.
(5) At the end of one batch the mixture is removed from the pipe by compressed
air which is fed into the system. Before standstill for longer periods the pipe
is cleaned by feeding in gravel or special torpedo-like cylindrical solids called
“Molche” (newts).
The most recent investigations concerning this transport system show that it
might be necessary to reduce the pellet-content to < 50 vol% in the fluid mixture.
This would cause an increase of volume of about 25% over that suggested above.
6.2. Ventilation system
The cavern ventilation system has two tasks:
(1) maintenance of defined pressure conditions close to atmospheric
pressure in the free space of the cavern and
(2) decontamination of airborne radioactivity from the off-gas system.
IAE A-SM-243/16 381
FIG.5. Maximum temperature of the reference product in the cavern plotted against time.
Airborne radioactivity is generated by the circulation of the cavern air, which
is contaminated during the filling procedure either with aerosols or with НТО
containing vapour. The required decontamination factor is at maximum 70, which
is technically feasible.
The main feature of the proposed cavern ventilation is a vented air circulation
system restricted to the inlet area at the top of the cavern in order to avoid major
exchange of the vented air with the contaminated cavern air.
The following factors are relevant for the layout of the cavern ventilation
system :
( 1) atmospheric pressure gradient (max. ± 15 mb/h);
(2) air expansion due to hydration heat release;
(3) compensation of waste volume.
7. SAFETY ANALYSIS
7.1. Heat production
The temperature of the solidified waste in the cavern results from contributions
from the heat of hydration, the decay heat of the fission products and the plutonium,
and the temperature of the surrounding salt (40°C).
382 KÔSTER et al.
Calculations have shown that the maximum temperature in the cavern is
mainly determined by the contribution of the heat of hydration during setting of
cement grout. The contribution of the fission products is small, compared with
the heat of hydration. The temperature rise due to the decay heat of plutonium
is negligible (about 2 K).
The calculated course of the temperature versus timecurve in the system
considered here (complete setting of the granules prior to introduction into the
cavern) is given in Fig.5. It can be noted that the temperature in the cavern becomes
lower with an increasing sand-to-cement ratio (S/C).
In the case of S/C = 2 (where about 60% of the furnace slag cement is
substituted by sand) and three operations per year, the maximum temperature in
the cavern is 85°C and thus below a temperature of 90°C, which is currently
assumed to be the upper limit because of product quality reasons.
Generally it is of advantage with respect to the heat generation in the cavern
to select a small thickness for the layers introduced (i.e. many operations per year)
and large time intervals between the operations in addition to a large sand-to-
cement ratio.
7.2. Stability of the cavern
Based on experience to date on underground cavities in saline formations,
the following statements can be made:
The excavation of large underground cavities in saline formations by solution
mining can be considered as technically advanced throughout the world. The
depths of such caverns are between 600 and 1500 m, the distance between the
axes are 150 and 250 m. The ceilings have thicknesses of at least 100 m. In the
Federal Republic of Germany there are about 40 large storage caverns in operation
with volumes generally between 2 and 5 X 10s m3. There are far fewer caverns
excavated conventionally, i.e. by drilling and blasting or mining techniques. This
type is represented by the prototype cavern Asse with a volume of 10 000 m3.
As the storage rooms used for the purposes considered here are operated at
atmospheric pressure only, a direct transfer of knowledge gained of gas or
pressurized-air storage caverns is not relevant.
The prototype cavern at Asse with atmospheric pressure conditions is the
subject of close in situ investigations into its tectonic behaviour.
Based on the experience gained from the prototype cavern after 15 months
of observation, it can be stated that it was absolutely stable during this period of
time; critical deformations were nowhere observed.
Definitive statements concerning the stability of caverns of 75 000 m3,
based on finite element calculations, will be worked out within the next two years.
IAEA-SM-243/16 383
ACKNOWLEDGEMENTS
The R &D work for this project is sponsored by the Federal Ministry of
Research and Development and jointly executed by the following institutions:
Amtliche Materialprüfanstalt für Steine und Erden, Clausthal-Zellerfeld (AMPA)
Gesellschaft für Strahlen- und Umweltforschung mbH, Wissenschaftliche
Abteilung, Clausthal-Zellerfeld (GSF)
Kernforschungszentrum Karlsruhe (KfK)
Gelsenberg AG, Essen (GAG)
F.J. Gattys Verfahrenstechnik GmbH, Neu-Isenburg (GATTYS)
Nukem GmbH, Hanau (NUKEM).
The authors belong to the project management and they would like to express
their thanks to the numerous scientists of the above institutions whose contri
butions were summarized in this paper.
REFERENCES
[1] RUDOLPH, G., KÔSTER, R., “ Verfestigung von LAW-, MAW- und Tritiumabwassern aus der Wiederaufarbeitung” , Reaktortagung Hannover 1978, Zentralstelle für Atom- kernenergie D.okumentation (1978) 482.
[2] RUDOLPH, G., KÔSTER, R., “ Immobilization of strontium and caesium in intermediate- level liquid wastes by solidification in cements” , Scientific basis for Nuclear Waste Management, Vol. 1, Plenum Publishing Corporation New York (1979) 467—470.
[3] WITTE, H.O., KÔSTER, R., “ Recent developments in low- and intermediate-level waste fixation by cement” , Ceramics in Nuclear Waste Management (Proc. Symp. Cincinnatti, 1979) (to be published).
DISCUSSION
J.K. WIERZCHOÑ: How many batches will you have in a period of five
years, and how will the free pipe be cleaned?
R. KÔSTER: We are thinking of 3—6 batches per year and this corresponds
to 15-30 batches for five years of operation. It is planned to carry out the pipe
cleaning with water and gravel and with torpedo-like cylindrical solids called
“Molche” (newts).
J.K. WIERZCHON: Do you really think that it is an easy job to cut the
free-hanging pipe?
R. KÔSTER: I don’t think it is an easy job but it is possible with the present
state of the art.
384 KOSTER et al.
H. RÛTHEMEYER: Am I right in assuming that the concept is of advantage
only in an integrated “Entsorgungszentrum”?
R. KOSTERYou are quite right. The integrated concept is the most
necessary boundary condition because this would avoid transporting liquid waste
on public roads.
A.-M.-L. BOULANGER: You are proposing to maintain under atmospheric
pressure for five years a cavern 30 m in diameter at a depth of 1000 m. What
temperature will you obtain in this cavern, and what stability problems do you
anticipate?
R. KOSTER: Owing to the hydration heat release in the cavern, we will get
temperatures of up to ~ 90°C (see Fig.5). Two years ago the project management
initiated calculations to determine the stability of caverns. The results based on
finite element calculations have so far been very poor, but we have identified the
problem and are working on it.
D.B. STEWART : Could you indicate the estimated density of the
consolidated product in situ? What temperature will it attain over time?
R. KÔSTER: The final product density will be 2 ± 0.2 g cm"3. The
temperature as a function of time is shown in Fig.5. In 50 years we shall have
decreasing temperatures, which will end up somewhere between 40° and 50°C.
M.K. PIMENOV: It appears that you intend to construct the cavern by
solution mining. It is known that to obtain 1 m3 of effective volume one needs
10—20 m3 of water. In this concept how do you ensure safe (from the environ
mental standpoint) removal or disposal of 350 000—370 000 m3 of salt water
which will be formed in the process?
R. KOSTER: The cavern can be constructed either by solution mining or by
dry mining techniques.
The discharge of salt solutions from solution mining has not so far been
checked in a precise manner because this “in situ concept” is still in the stage of
feasibility study. Up till now we have been thinking in terms of discharging some
quantities perhaps into the river Elbe (if salt contamination of the river remains
within permissible limits) and larger quantities via pipelines into the sea.
J. HAMSTRA: Figure 2 shows that the cavern filling takes place from a
mixing installation situated at the surface. Would it be feasible to do the filling
from inside the mine, too?
R. KÜSTER: During the first concept phase of the project we looked at a
wide variety of filling concepts (different cavern sizes, filling from inside the mine,
filling of liquid/solid wastes and so on). Mainly on the basis of transport criteria
(for example, minimized erosion of the pipe walls, dose rates to the staff) we
selected the mixing installation at the surface. Filling from inside the mine is
feasible in principle.
DISPOSAL IN DEEP GEOLOGICAL FORMATIONS:
CRYSTALLINE ROCKS
(Session V)
Chairman
P.A. WITHERSPOON
United States of America
IAEA-SM-243/68
PREMIERES EVALUATIONS DES POSSIBILITES
D’EVACUATION DES DECHETS RADIOACTIFS
DANS LES ROCHES CRISTALLINES
A. BARBREAU, Y. SOUSSELIER
CEA, Institut de protection et de sûreté nucléaire,
Centre d’études nucléaires de Fontenay-aux-Roses,
Fontenay-aux-Roses
M. BONNET, J. MARGAT, P. PEAUDECERF
Bureau de recherches géologiques et minières,
Service géologique national, Orléans
P. GOBLET, E. LEDOUX, G. de MARSILY
Ecole nationale supérieure des mines de Paris,
Centre d’informatique géologique, Paris,
France
Abstract-Résumé
A FIRST EVALUATION OF POSSIBILITIES OF RADIOACTIVE WASTE DISPOSAL IN CRYSTALLINE ROCKS.
The French research programme on radioactive waste disposal in deep geological formations is focused primarily on crystalline formations, under a contract financed jointly with the Commission of the European Communities. Although theoretical and experimental work is still in progress, it is proposed that an initial evaluation of the effectiveness of confinement in such a waste repository should be made using simulations which, although simplified, are nevertheless realistic since they are based on the results of the first sets of measurements in the field. Even if this approach does not lead to any definitive conclusions about the “ safety” of the geological barrier, it will enable the role of each of the elementary mechanisms and associated parameters to be appreciated; and these results will make it possible to define more precisely the work which still needs to be done.
PREMIERES EVALUATIONS DES POSSIBILITES D’EVACUATION DES DECHETS RADIOACTIFS DANS LES ROCHES CRISTALLINES.
Le programme d’étude français concernant les possibilités d’évacuation des déchets radioactifs dans les formations géologiques profondes est en priorité consacré à l’étude des formations cristallines dans le cadre d’un contrat à frais partagés avec la Commission des Communautés européennes. Bien que les travaux théoriques et expérimentaux soient actuellement en cours, il est proposé de procéder à une première évaluation de l’efficacité du confinement d’un dépôt à l’aide de simulations simplifiées mais réalistes car fondées sur les résultats des premières campagnes de mesures sur le terrain. S’il n’est pas possible d’en tirer des conclusion définitives sur la isûreté) de la barrière géologique, cette approche permet d’apprécier le rôle de chacun des mécanismes élémentaires et des paramètres associés, ces résultats conduisant à une meilleure définition des travaux qui restent à réaliser.
387
388
AVANT-PROPOS
BARBREAU et al.
Depuis plusieurs années, la France a entrepris d'étudier
la possib ilité d'éliminer les déchets radioactifs dans les forma
tions géologiques, et plus particulièrement dans des terrains
granitiques. Ce programme d'études, réalisé par le Commissariat à
l'Energie Atomique, le Bureau de Recherches Géologiques et Miniè
res et l'Ecole Nationale Supérieure des Mines de Paris, se dérou
le dans le cadre d'un contrat à frais partagés avec la Commission
des Communautés Européennes. Il comprend la recherche de massifs
granitiques favorables, l'étude de leurs caractéristiques géolo
giques et hydrogéologiques ainsi que des études associées dont
l'o b je c tif est d'évaluer la capacité de confinement de la barriè
re géologique.
Au stade actuel, les études consistent à qua lifie r ce
type de roche grâce à l'iden tifica tion et à la quantification des
différents paramètres qui permettent de modéliser le transfert de
la radioactivité.
Par a illeurs , aucune décision n'a encore été prise en
France quant au choix final d'une formation géologique particu liè
re. Ce choix devrait dépendre des résultats des études en cours.
INTRODUCTION
L'enfouissement1 des déchets radioactifs dans des for
mations cristallines profondes est étudié ic i sous l'angle des
éventualités de retour ultérieur de la radioactivité vers l'e nv i
ronnement, c'est-à-dire des risques d'impact sur la biosphère.
1 C’est-à-dire l’évacuation définitive, après une phase temporaire de stockage avec récupération possible.
IAEA-SM-243/68 389
Les roches cristallines étant, comme toute roche, suscep
tibles de contenir et de permettre le déplacement de faibles quan
tités d'eau, le confinement des déchets repose essentiellement sur
trois facteurs de sûreté :
- les mécanismes de mise en solution du déchet et la tenue des bar
rières technologiques a r t if ic ie lle s qui l'entourent ;
- la possib ilité et les modalités des déplacements d'eau et de mi
gration d'éléments en solution dans la formation cris ta lline dans
son ensemble ;
- la conception et la technologie du dépôt lui-même et de ses voies
d'accès.
Les résultats qui suivent portent exclusivement sur le
deuxième point ci-dessus. En outre, dans cette première approche,
les hypothèses suivantes ont été adoptées :
- les circulations d'eau dans un massif granitique faiblement f is
suré sont considérées comme engendrées exclusivement par les con
ditions hydrologiques de surface à l'exclusion de tout effet ther
mique (naturel ou provoqué par l'introduction des déchets);
- les vitesses de migration des éléments passant en solution par
lix iv ia tion sont évaluées à partir de la vitesse de l'eau définie
dans les conditions ci-dessus et en prenant en compte la disper
sion cinématique et les différents phénomènes d'interaction avec
le milieu c r is ta llin : diffusion dans la matrice rocheuse à par
t i r des fissures, rétentions physico-chimiques aux contacts
solides-liquide.
Notre approche est basée sur l'emploi de modèles capables
de prendre en compte les hypothèses et les phénomènes ci-dessus et
consiste principalement en une étude de sensib ilité aux divers pa
ramètres qui interviennent dans la constitution de ces modèles. Les
gammes de variation admises pour ces paramètres ont été autant que
390 BARBREAU et al.
possible choisies en fonction des premiers résultats expérimentaux
acquis sur le terrain. On a cherché à déterminer quels sont les
paramètres qui ont la plus grande influence sur les conclusions en
termes de sûreté afin d'en déduire les points sur lesquels on doit
faire porter l 'e ffo r t dans la suite du programme de terrain qui
prévoit notamment des investigations à plus grande profondeur.
1. DEPLACEMENTS NATURELS DES EAUX
Il est particulièrement important de prévoir la vitesse
de déplacement du vecteur éventuel que constitue l'eau souterrai
ne. Pour mieux préciser ces déplacements, on a réalisé des simula
tions sur un modèle hydrodynamique d'écoulement. Dans cette pre
mière évaluation, nous n'avons pas pris en compte la dispersion
cinématique qui n'a pas d'influence sur la vitesse moyenne de l 'é
coulement. Son introduction ne présenterait d 'a illeurs aucune d if
ficu lté .
1.1. Description du modèle de l'écoulement souterrain
La simulation numérique porte sur les écoulements se ma
nifestant dans un massif en situation d 'interfluve. Ce bloc est
délimité par des parois verticales, un fond horizontal à 2 ООО m
et la surface du sol. Les rivières sont les seuls exutoires du
système. Par raison de symétrie, les simulations peuvent être ré
duites à la moitié de l'in terfluve . On néglige les phénomènes ther
miques. Ainsi, le mouvement des eaux souterraines est provoqué
uniquement par les conditions hydrographiques et morphologiques de
surface qui fixent les conditions aux limites amont (apport) et
aval (émergence) (cf. figure 2).
Dans une précédente communication (BONNET et a l . , réf. [1])
i l a été montré qu'une te lle approche permet en effet de reproduire
avec une bonne approximation les "écoulements souterrains" consta
tés dans un bassin hydrologique jaugé, situé en massif granitique.
IAEA-SM-243/68 391
1.1.1. Hypothèses_et_condi t i ons_aux_]i mi tes
A l'échelle macroscopique "régionale" considérée, les hy
pothèses suivantes ont été retenues :
- les écoulements sont bi-dimensionnels dans un plan vertical et
obéissent à la lo i de Darcy; en régime permanent les débits obte
nus sont donc proportionnels aux perméabilités pour une réparti
tion donnée de la charge hydraulique ;
- le milieu est supposé continu et de caractéristiques isotropes;
les zones de discontinuité telles que les fractures n'ont pas
été représentées en elles-mêmes, mais seulement par leur effet
sur la perméabilité moyenne d'un volume de terrain à la dimen
sion d'une maille ;
- le milieu est "régulièrement" hétérogène : la perméabilité à la
surface étant Ks, on impose différentes lois de décroissance
avec la profondeur z.
On a choisi des lo is théoriques de type exponentiel et aussi des
lois semi-empiriques d'après CARLSSON et OLSSON (réf. [2])et
SNOW (réf. [3]) (cf. figure 1) :
a) Kz = Ks lO-z/SOO
b) Kz = Ks , 10-Z/300
c) Kz = Ks îo-z/loo
d) Kz = Ks z_2»5 (Carlsson et Olsson)
e) Kz = Ks z"1»6 (Snow)
Ces lo is ont été introduites en considérant différentes
valeurs de la perméabilité proche de la surface, Ks , celles-ci
devant rester dans l'in te rva lle qui nous paratt le plus probable
d'après nos investigations de terrain : soit entre 10"4 et
10"6 m/s (à l'exception de la lo i dite de Carlsson et Olsson
où Ks = 3.10-3 m/s).
392 BARBREAU et al.
Ю-" Ю '7
2 0 0 0 -
t ( m) .
A / h /
‘ t ¡ / t f /1 H 1
Á V i
/ Л /Ч У! j /
о 2 — • —
b 2 . ^ o —
c 2 — ù —
# 3 ----- û - —
I (0 •(</•0048,1)
i (Q - U / * 0 0 * M )
, Ю -< */ » o o + « >
, Ю -(i/ioo ♦ о , io - <« /юо**)■ io -<*,•»•« i *M>, ,o -<*,«»•• • «■»»i |Q >*« i > i , t >
> Ю - t 'и* ■* «1 » Ю -< 1
[ C A K L S S O N ond О С Э З С н ]
[s n o w ]
FIG.L Lois de variation des perméabilités en fonction de la profondeur.
Ю -* K * ( m )
ptrm éobi l it * *
IAEA-SM-243/68 393
h - ■ 4
|рШ |Ш
i I I m p i r m • o b i # i f
@ @ ®
FIG.2. Représentation schématique du modèle numérique d’écoulement. Conditions aux limites et exemple de lignes de courant.
394 BARBREAU et al.
Comme le montre la figure 2, les conditions aux limites
suivantes ont été introduites :
Limite supérieure : celle-ci est supposée à potentiel imposé :
d'une part l'exutoire unique est représenté par l'intersection
d'un cours d'eau à niveau constant, d'autre part la surface piézo-
métrique libre est aussi supposée constante au cours du temps
déterminée par la surface topographique comme on l'observe le plus
souvent dans les bassins hydrologiques granitiques. Ceci équivaut
à admettre comme hypothèse simplificatrice que le flux d'apport est
constant. Dans les conditions réelles, ce n'est naturellement pas
le cas, mais les potentiels sont peu variables.
On note ДН l'écart des charges entre le maximum piézo-
métrique et l'exutoire (minimum).
En outre, pour tenir compte des conditions hydrologi
ques régionales, on a lim ité les flux in filtrés à 5 1/s.km2, soit
150 mm d'eau par an environ ; ce qui nous a conduit,pour la plu
part des lois de variation de la perméabilité, à considérer Une
valeur maximale pour la perméabilité de surface Ks.
Une augmentation de ce débit d'apport, par exemple pour
se rapprocher d'autres conditions locales, modifierait d'autant
les vitesses de déplacement.
Limite inférieure : à 2 000 m de profondeur, supposée imperméable.
Limites latérales à flux nul assimilées à des limites imperméables:
suivant les cas, la zone simulée a une largeur L = 1 250 m ou 750 m.
1.1.2. Modèle_utilisé__et_méthode_._de_calcul
Le programme de calcul u tilis é détermine les charges
et les débits par une méthode aux différences finies classique
(surrelaxation). Le domaine d'étude a donc été discrétisé en "cou
ches", chaque couche étant elle-même redécoupée dans le plan ho
rizontal. Les pas de discrétisation dans les deux directions ne
sont pas nécessairement uniformes (fig . 2).
Dans le cas présent, un découpage en mailles rectangu
laires a été adopté : 250 à 1 000 m dans le plan horizontal et
verticalement la discrétisation a été choisie en fonction de la
topographie et de la lo i de décroissance de la perméabilité.
Le bloc étudié a donc été découpé en 20 couches environ de 5m à
800 m d'épaisseur, ce qui correspond à une centaine de mailles
utiles.
Enfin, le programme permet de tracer des cartes pié-
zométriques en coupe verticale ainsi que les lignes de courant.
En adoptant une valeur de porosité cinématique on peut également
calculer le temps mis par l'eau pour atteindre l'exutoire à par
t ir de n'importe quel point du domaine.
IAEA-SM-243/68 395
1.2. Temps de transfert de l'eau. Influence des paramètres
Les points de départ de l'eau ont été supposés â des
profondeurs et des positions différentes (cf. fig . 2) :
a à la verticale de l'exutoire
e en position moyenne (entre a et y)
y à la verticale du dôme piézomêtrique,
les temps de parcours correspondants étant notés ta ' t^ et t^ .
Les résultats suivants ont été obtenus en prenant com
me porosité cinématique une valeur moyenne que nous avons mesurée
dans un massif granitique réel,2.10-4 (CALMELS et a l, réf. [4]),
396 BARBREAU et al.
et correspondant évidemment à une porosité de fissure. Avec une
autre valeur ш, tous les temps de transfert qui suivent devraient
être multipliés par le rapport w/2.10“4.
1.2.1. Infly§nçe_de_]a_loi_de_variation_des_perméabi-
H tés.
La sensib ilité des temps de déplacement à la réparti
tion des perméabilités est très importante. Ainsi par exemple,
pour ДН = 25 m et L = 1 250 m, so it un gradient de 2 %, les
durées de parcours en années à partir de 1 000 m de profondeur
sont indiquées au tableau I. De même, les durées de parcours en années depuis
1 500 m de profondeur jusqu’à l’exutoire superficiel sont données au tableau II,
pour les mêmes lois.
Ainsi, on constate une très forte sensib ilité des durées
de déplacement à la loi de variation des perméabilités avec la pro
fondeur , les durées de déplacement de l'eau pour des points situés
à 1 500 m de profondeur dans le plan médian (tg)par exemple étant
comprises entre 700 ans et 40 000 années.
1.2.2. Inflyençe_du_gradient_régional
Le gradient moteur ДН/L dépend des conditions hydro
graphiques superficielles ; ДН est lié à la topographie et 2L est
l'écartement moyen entre deux cours d'eau.
- Si l'écartement L est fixé , les vitesses sont directement pro
portionnelles à ДН dont les valeurs sont connues et peu varia
bles pour un massif granitique donné.
- Par contre, si la charge ДН est constante, l'influence de l'écar-
tement des axes de drainage (traduit par la distance L) est com-
IAEA-SM-243/68 ' 397
TABLEAU I. DUREE DE PARCOURS DEPUIS 1000 m
l o i s
"a" Kz * 10"6.10_z/ 500
"b11 Kz = 3,2.10-6.10"z/30°
"c" Kz = 10-5-10-z/ 100
CL Kz = 3,2.10-3.z-2’5
"e"l Kz = 3,2.10“4.z " b 6
"eu2 Kz = ÎO-5.z-1.6
4000 (m/s)
1 0 " 8
1.5.10-11
non fissur
10-10
5.10"9
1.6. 10- 1°
t a(ans) tg(ans) y ans)
30 38 600
120 220 2 ,1 .104
temps très grands non calculés
4 ,4 .103 5 ,6 .103 7.6.104
130 140 1,6.103
4.103 4.5.103 5.2.104
TABLEAU II. DUREE DE PARCOURS DEPUIS 1500 m
lois K1500(m/ S) ta (ans) t B(ans) ty(ans)
"a" 10"9 530 700 7.8.Ю3
"b" 3.10-11 2,3.104 3,5.104 1,7 .105
"d" 3,6.10-H 3,6 .104 4 ,6 .104 2.2.105
"e"l 2 ,6 .10"9 720 950 3 ,7 .103
"e"2 8,3.10“11 2.3.104
OСО 1,2 .105
plexe et varie selon la lo i de perméabilité et le point considé
ré. A insi, quand L est multiplié par 0,6 les durées de déplace
ment sont multipliées par les facteurs suivants :
Depuis 1 000 m facteurs pour t a *3 tY
Loi "a" 1.2 1,3 0,15
"b" 1,2 1,0 0,52
11 d" 1,0 1,1 0,8к e «
1,1 1,3 0,8
398 BARBREAU et al.
Depuis 1 500 m facteurs pour ta S tУ
Loi "a" 2,6 2,1 0,41
"b" 1,3 1.2 1,8
"d" 1,6 1.9 1,8
"e" 1,6 1.5 1,6
On constate d'une part que pour les points de départ
profonds que nous avons considérés (1000 et 1500m) une plus gran
de densité du réseau hydrographique superficiel, même si elle
conduit à des gradients piézométriques de surface plus importants,
n'entraîne pas des vitesses de circulation plus grandes. Leur
ordre de grandeur est conservé dans la gamme possible de varia
tion des gradients дН/L.
En définitive, les conditions hydrographiques régiona
les influent peu sur la sûreté du site.
1.2.3. Influ§nçe_de_1a_gorosité_çinématigue
Rappelons que les temps de parcours sont proportionnels
à la porosité cinématique considérée et que les cas précédemment
étudiés sont relatifs à une porosité de fissure (2.10"^).
Ces coefficients de porosité ne subiront de variation
importante que s'ils intègrent les effets des rétentions sur les
parois des fissures ou même dans la masse de la roche conformément
à la notion de porosité apparente évoquée dans les paragraphes
suivants.
2. INFLUENCE DES DIFFERENTS PHENOMENES DE RETENTION
Les résultats qui viennent d'être présentés supposent
que la migration du polluant s'effectue exclusivement suivant le
mouvement moyen de l'eau dans les fissures du granite auxquelles
IAEA-SM-243/68 399
est attribuée la porosité cinématique. Ils ne prennent donc pas
en compte les différents mécanismes des rétentions qui peuvent
intervenir au cours de la circulation des solutions. Nous allons
maintenant tenter une quantification de ces effets.
Afin de mettre clairement en évidence les effets des
différents mécanismes de rétention envisagés, on a choisi volon
tairement des hypothèses simplificatrices concernant l'hydrody
namique des écoulements. C'est pourquoi les valeurs numériques
obtenues dans cette seconde partie de l'étude ne devront pas être
directement comparées à celles de la première partie.
2.1. MODELISATION DES PHENOMENES DE RETENTION
Il n'est pas question ici de décrire en détail l'en
semble des mécanismes physico-chimiques responsables de la réten
tion des radioéléments dans la roche. On recherchera simplement
une représentation globale compatible avec l'approche en "milieu
continu" adoptée pour la modélisation des écoulements, dans le
cas des roches à faible perméabilité (de fissure).
On. distinguera deux types de phénomènes d'interaction
entre la solution et la roche :
- un phénomène lié à la circulation de l'eau dans les fractures
traduisant l'interaction entre les solutés et les minéraux tapis
sant les fissures; cet effet sera fonction de la surface de
fracture léchée par l'eau et de son aptitude à la fixation des
radioéléments;
- un phénomène lié à la pénétration des polluants dans la matrice
rocheuse grâce à la diffusion dans l'eau occupant les vides
correspondant à la porosité d'interstices du granite à partir
400 BARBREAU et al.
des fractures; cet effet sera fonction du volume des vides pré
sents dans la matrice, du coefficient de diffusion des substan
ces dissoutes ainsi que de la faculté de rétention des minéraux
constituant le granite.
Cette approche revient à décomposer les vides saturés
d'eau du milieu géologique en deux fractions, comme l'ont fait
NORTON et KNAPP (réf. [5]):
- une fraction offerte â l'écoulement ou "porosité cinématique",
- une fraction où la circulation du fluide est négligeable, mais
permettant la diffusion des substances dissoutes, ou "porosité
diffusionnelle".
Nous supposerons sans action sur la migration des ra
dioéléments la porosité résiduelle constituée par les pores non
connectés entre eux, ainsi que la diffusion dans les solides.
2.2. MISE EN EQUATIONS
2.2.1. Retention_dans_les_fissures
Nous admettrons tout d'abord que le déplacement de l'eau
dans les fissures s'effectue dans un milieu continu possédant les
caractéristiques hydrauliques régionales du massif granitique.
Compte tenu de la lenteur des circulations attendues dans un mas
sif susceptible de constituer un site d'évacuation de déchets, nous
supposerons qu'il existe à chaque instant un équilibre entre la
quantité de matière présente en solution et celle retenue sur la
paroi des fissures.
Ce type d'interaction sera alors représenté par une loi
de rétention linéaire et instantanée de la forme :
F = Kd C
IAEA-SM-243/68 401
ой C est la concentration dans l'eau en mouvement dans les fissu
res [M] [L]"3
F est la masse fixée par unité de surface [M][L]
Kd est le coefficient de distribution dans les fractures [L].
De récentes expériences de traçage effectuées sur un
massif granitique fissuré français ont permis une estimation de
la valeur de pour le strontium et le césium (réf. [4] et [6]):
Strontium : 10"* m
Césium : 5.10” m
2.2.2. Diffusion_dans_la_matriçe_granitigue
Ce mécanisme sera représenté en imaginant un second
milieu continu superposé au premier et possédant les caractéris
tiques du granite massif, dans lequel le mouvement de filtration
de l'eau est négligeable et ой seule la diffusion moléculaire des
polluants à partir du réseau des fissures est possible.
Cette diffusion sera supposée s'effectuer suivant la
loi de FICK :
Ф = -d ш‘ grad C '
ой ф est la densité de flux de diffusion
C' la concentration dans l'eau de la matrice rocheuse [M][L]
ш' la porosité diffusionnelle
d le coefficient de diffusion en milieu poreux [L]2 [T]-1.
d peut être estimé à partir du coefficient de diffusion moléculai
re en eau libre do (réf.[7])
402 BARBREAU et al.
do varie peu avec les différentes substances et est de l'ordre de
10-9 m2 /s â 20°C.
Compte tenu de la lenteur de cette diffusion, nous ad
mettrons également une loi de rétention linéaire et instantanée
sur les grains de la matrice de la forme :
F' = K'd C'
où F1 est la masse de polluant retenue par unité de masse de mi
lieux poreux
et K'd le coefficient de distribution en milieu poreux du granite.
2.2.3. §9yation_du_transfert
La présence de polluant dans le massif sera alors dé
crite par deux concentrations :
C dans l'eau contenue dans les fissures (dans la porosité cinéma
tique
C' dans l'eau contenue dans la matrice (dans la porosité diffusion-
nelle.
Les équations de transfert dans chacun des milieux con
tinus s'écrivent :
яГdiv (D grad C) - div (VC) = (ш + y Kd)-^ + уф dans les fissures
div (ai' d grad С') =Гш‘ + (1 - o>')p K'd] — dans la matriceL -J at
avec la relation Ф = - dio' grad C' qui assure le couplage entre les
deux milieux.
Les notations supplémentaires introduites sont :
D coefficient de dispersion en milieu fissuré [L]2 [T]-*
V vitesse de DARCY de l'écoulement [L] [T]-1
u porosité cinématique
y surface spécifique de fracture [L]"^
p masse volumique de la roche sèche.
Remarquons qu'avec les hypothèses faites sur les lois de rétention,
tout se passe comme si le milieu fissuré possédait une porosité
apparente :
ша = a) + p Kd
et la matrice, une porosité apparente :
(ü 'a = ш' + (1 - u ) ' ) p K'd
De plus si la vitesse de diffusion dans la matrice peut être con
sidérée comme infinie, la concentration C devient à chaque ins
tant égale à C' et le système se réduit à une équation unique :
div (D grad C) - div (VC) = |\o + u' + y Kd + (1-ш')р K'd]|£
et tout se passe comme si Гоп avait un milieu continu unique de
porosité apparente :
ш"а = ш + ш' + y Kd + (l-(o')p K'd
2.3. EXEMPLES D'APPLICATIONS
Afin de faire apparaître le rôle des divers "mécanis
mes de retard" possibles, nous avons appliqué le modèle qui vient
d'être décrit en faisant varier les paramètres correspondants selon
IAE A-SM-243/68 4 03
404 BARBREAU et al.
la même approche que pour les phénomènes de transfert de l'eau.
Mais pour simplifier, nous avons supposé les caractéristiques du
milieu homogènes dans l'espace et dans le temps et considéré un
écoulement unidirectionnel pouvant être grossièrement assimilé au
développement d'une ligne de courant issue d'une des simulations
hydrodynamiques présentées ci-dessus et à laquelle correspondraient
les paramètres suivants :
. longueur du transfert : 1000 m
. gradient hydraulique : 0,005
. perméabilité de fissures : 10-8 m/s
. porosité cinématique : 10-4
. dispersivité : 20 m
. surface spécifique de fracture : 0,1 m~l (ce qui correspond par
exemple à une famille de fractures parallèles planes distantes
entre elles de 10 m)
. coefficient de diffusion dans l'eau : 10"^ m2/s.
A. titre indicatif, le jeu de paramètres choisi se
rapprocherait des conditions qui ont conduit aux temps de parcours
les plus rapides lors des simulations hydrodynamiques de la pre
mière partie.
Nous simulerons alors la migration d'une pollution
constante égale à l'unité à partir de l'instant initial zéro,
sans tenir compte de la décroissance radioactive.
Nous ferons.au cours des calculs, varier les paramè
tres caractérisant la rétention Kd et K'd.
Le système d'équations donnant la concentration C de
l'eau dans les fissures a été résolu numériquement dans les 3 cas
suivants :
IAE A-SM-24 3/68 405
10 100 1000 10 ООО 100 ОООtemps en années
FIG.3. Concentration à l ’exutoire du massif. Influence de la porosité diffusionnelle
(Kd = 0; K'd = 0).
C / t o u r c e
1000 10 ООО 100 ООО 1 000 000temps en années
FIG.4. Concentration à l ’exutoire du massif Influence de la porosité diffusionnelle
(Kd = 10~1 m, K ’d = 0).
1er cas : pas de rétention dans les fissures ni dans la matrice
(Kd = 0, K'd = 0). Quatre simulations ont été réalisées avec les
valeurs suivantes pour la porosité diffusionnelle (fig. 3) :
ш'а = 0 - 10"4 - 10-3 et 10-2
406 BARBREAU et al.
1000 10 ООО 100 ООО 1 ООО ОООtemps en années
FIG.5. Concentration à l'exutoire du massif. Influence de la rétention dans la matrice
(Kd = 10 '1 m; œ’ = 10~3).
2ême cas : rétention dans les fissures seulement (Kd = 10"! m,
K'd = 0). Nous avons adopté pour Kd la valeur obtenue lors des
expériences in situ sur granite pour le strontium, soit Kd = 10_1m,
et l'on a utilisé les mêmes valeurs que dans le premier cas pour
la porosité diffusionnelle (cf. fig. 4).
Cette valeur de Kd est utilisée à titre de démonstration sans pré
tendre être représentative de la valeur réelle en profondeur dont
la détermination devra être faite pour chaque cas particulier.
3ëme cas : rétention dans les fissures et dans la matrice. On a
conservé la valeur Kd = 10~1 m pour le coefficient de distribution
dans les fissures et l'on a adopté une porosité diffusionnelle
moyenne de W 3. On a ensuite effectué des simulations pour les
valeurs 0 - 0,1 - 1 - 10 et 100 ml/g du coefficient de distri
bution en milieu poreux (cf. fig. 5).
IAEA-SM-243/68 407
Les résultats obtenus appellent les remarques suivantes:
- la durée de migration des éléments reste toujours gouvernée par
les fissures, la matrice n'ayant qu'un effet d'amortissement de
la concentration (effet qui n'est pas à négliger dans la prati
que) ;
- le rôle de la matrice varie d'ailleurs en relation avec le com
portement dans les fissures : l'effet retardateur de la diffu
sion est d'autant plus grand que le déplacement de l'eau dans
les fissures est lent ;
- l'influence de la matrice varie notablement avec la porosité
diffusionnelle ; toutefois cette influence ne devient sensible
que pour des valeurs élevées de celle -ci (de l'ordre de 10“)
peu vraisemblables dans un granite ;
- pour un écoulement donné, la courbe de restitution (cf. fig. 5)
tend vers une position limite lorsque K'd augmente; ce fait doit
être attribué à l'influence limitative de la vitesse de diffu
sion dans la matrice par rapport à sa capacité de rétention.
CONCLUSIONS
L'étude générale de l'aptitude des massifs granitiques
à servir de réceptacle à des dépôts de déchets radioactifs effec
tuée en France, a comporté dans sa partie théorique un essai de
recensement de tous les mécanismes physiques qui devraient condi
tionner la sûreté d'un site dans ce type de formation.
Dans l'étude de la sûreté de l'évacuation en formation
géologique, nous n'avons considéré en première analyse que les
plus influents de ces mécanismes en les schématisant volontairement.
408 BARBREAU et al.
Il devient alors possible d'utiliser des modèles simples mais vrai,
semblables parce que basés sur des résultats expérimentaux issus
d'essais sur le terrain ou en laboratoire.
On a cherché, à l'aide de ces modèles, à apprécier la
sensibilité des variables qui définissent le mieux la sûreté d'un
site (notamment le temps de migration depuis le site jusqu'à
l'environnement de surface) en fonction des divers paramètres du
milieu confinant.
Certes, une telle étude paramétrique ne permet pas en
core, compte tenu de la trop grande incertitude qui règne sur les
valeurs réelles des paramètres considérés, d'évaluer le degré de
sûreté qu'offrirait un dépôt de déchet dans un massif granitique.
Elle nous a permis cependant d'aboutir à quelques ré
sultats généraux importants :
- la répartition régionale des perméabilités au sein du massif
cristallin a une forte influence sur la durée de déplacement
de l'eau : les évaluations plausibles que l'on pourrait en don
ner a priori donnent en effet des résultats très dispersés; il
apparaît donc indispensable de la connaître dans le massif con
cerné pour apprécier l'efficacité du confinement du dépôt;
- la rétention éventuellement exercée par la roche met en cause
deux facteurs :
. un facteur statique lié au coefficient de distribution dans
la matrice ;
. un facteur cinétique lié à la porosité diffusionnelle et à la
vitesse des transferts dans les fissures.
IAEA-SM-243/68 409
Le second facteur pouvant occulter fortement le premier, il ap
paraît nécessaire de les étudier dans les conditions locales du
dépôt.
Par conséquent, la présente étude paramétrique réalisée
servira de guide pour conduire les travaux de terrain et de labora
toire qui devront comprendre :
- la réalisation de forages profonds pour apprécier les lois de
variations des paramètres (hydrodynamiques notamment) du milieu
avec la profondeur ;
- des essais de laboratoire, confrontés-en permanence avec des
essais in situ, pour préciser l'importance de certains mécanis
mes relatifs aux échanges entre radioéléments et le milieu
granitique (fissure et matrice).
Ainsi, compte tenu de la sensibilité du confinement à
la valeur des paramètres du milieu, il conviendra pour le choix
final de déterminer les paramètres avec le plus grand soin.
Références
[1] BONNET (M), NOYER (M.-L.), VAUBOURG (P) 1979 - Estimationdes vitesses des écoulements souterrains régionaux en milieu cristallin au moyen d'un modèle de simulation. SêmincuAz бил. la migration dea tiadionucléideA à t опдиг v-te dañó ta дёобркелг, ofiganibé. рал Ы Comm¿¿>A¿on dej¡ Communauté* Еилор&тпгл et l'OCVE Agence de Z'EneAgle. nucIêcUAe. В>шхг1Лел, 29-37 janvluA.
[2] CARLSSON (A.)» OLSSON (T.) 1977. Perméabilitétens variation i det svenska urberget. -SveA¿ge¿ Gzologlbka undeA 4dkiUng.
410 BARBREAU et al.
[31 SNOW (D.T.) 1968 - Hydraulic character of fractured metamor- phic rocks of the frontrange and implications to the Rocky Mountain Arsenal wells. - Quarterly oh the Colorado School oh M¿пел, vol. 63, n° 1.
[4] CALMELS (P.), GUIZERIX (J.). PEAUDECERF (P.), ROCHON (J.)1979 - Détermination des conditions de transfert de produits radioactifs dans un massif granitique fissuré au moyen d'essais ln s itu et d'essais sur échantillons. Séminaire ¿ил la migration des radlonucMld&> à longue, vie dam la. géosphere, organisé рал. la Commission des Communautés Européennes et l'OCVE Agence de V Energie nucléaire. Bruxelles, 29-37 janvier.
[5] NORTON (D.), KNAPP (R.) 1977 - Transport phenomena in hydro-thermal systems : the nature of porosity. -Amer. Journ. oh Science, vol. 277, o c t., pp. 913-936
[6] GOBLET (P.), LEDOUX (E.), MARSILY (G.)de, BARBREAU (A.)1979 - Représentation sur modèle de la migration des radioéléments dans les roches fissurées.Séminaire бит la migration des radionucZéldes à longue vie dam la géospkère, organisé par la Commission des Comunautés Européennes et l ’OCDE Agence de V Energie nucléaire. Bruxelles, 29-31 janvier.
[7] WYLLIE (M.R.J.), GREGORY (A.R.) 1953. - Trans. AIME, 198,p. 103
DISCUSSION
Y. INOUE: How did you allow for heterogeneity of formation in your
fundamental equation?
G.E. LEDOUX: The migration model in its theoretical application shown
here considers the overall behaviour of the geologic formation and therefore
assumes it to be homogeneous. We note, however, that the approach using two
continuous media (fissured medium and matrix) is one way of representing a
certain basic heterogeneity of granites. Moreover, the part of the paper dealing
with the influence of the permeability law on flow is intended precisely to
emphasize one aspect of heterogeneity.
IAE A-SM-24 3/68 411
Y. INOUE: What is the definition of K'd?
G.E. LEDOUX: K'd is a distribution coefficient expressed as the ratio
between the quantity of matter retained over unit mass of matrix and the quantity
of matter dissolved in unit volume of water.
J. HAMSTRA: Is it correct to interpret the result of your model calculation
in such a way that one may expect the retention capacity of a granite host rock
to be limited to 5000 years for about 20% of your source term and to 10 000 years
for about 80% of that term?
G.E. LEDOUX: No. As we have tried to show in our paper, the time for the
migration of radionuclides towards the environment through the barrier
constituted by granite depends on several factors, including especially the
variation of permeability in the rock mass where the flow takes place and also on
the retention capacity of the medium. Here we have confined ourselves to only
one example of purely theoretical application where a uniform average
permeability of 10-8 m/s and a relatively low Kd are considered. In this case, we
actually get a migration time of 10 000 years for 80% of the source term. Other
possible parameters could have been chosen and would have given other results.
It is quite obvious that in the case of disposal under much more favourable
conditions the migration time can be much longer, for example, of the order of
a million years. The purpose of our in situ studies is precisely that of investigating
formations which have all the characteristics conducive to the greatest containment
of radioactivity.
A. BRANDSTETTER: Do you plan to validate your model and if so, how
do you plan to obtain various parameters which distinguish porous media from
fractured media, for example, porosity and sorption coefficients for fractures?
G.E. LEDOUX: The models have already been validated in the course of
laboratory and in situ experiments carried out to a depth of 100 m. No experiment
has yet been performed in granites under the probable conditions of the disposal
site. The results presented here are based on values of plausible parameters,
due allowance having been made for current data. The safety study of a particular
site will obviously require careful determination of parameters which characterize it.
IAEA-SM-243/168
CANADIAN GEOSCIENCE RESEARCH AND
DESIGN CONCEPTS FOR DISPOSAL OF
HIGH-LEVEL WASTE IN IGNEOUS ROCKS
J.S. SCOTT
Geological Survey of Canada,
Ottawa
R.G. CHARLWOOD
Acres Consulting Services,
Niagara Falls, Ontario,
Canada
Abstract
CANADIAN GEOSCIENCE RESEARCH AND DESIGN CONCEPTS FOR DISPOSAL OF
HIGH-LEVEL WASTE IN IGNEOUS ROCKS.
The Canadian geoscience research programme, described in Part I of this paper, was
initiated in 1973 and has evolved into an integrated field and laboratory programme directed
toward verification of the concept of geological disposal of high-level radioactive waste in
igneous rocks. Granite, anorthosite, syenite and gabbro in the Precambrian Shield of Ontario
are the principal rock types o f interest. Geological, Geophysical, Rock Properties and Hydro-
geological Activities constitute the major groupings of tasks within the programme. Pending
public acceptance of field research, present field studies are confined to granitic rocks under
lying Atomic Energy of Canada Limited research establishments at Chalk River, Ontario, and
Pinawa, Manitoba. Tentative correlations have been established between hydraulic conductivity
and fracture intersects obtained from field data and laboratory measurements of formation
factor. Airborne gradiometer surveys provide a rapid means for determination of structural
discontinuities. Part II o f this paper covers design concepts for deep underground disposal in
igneous rocks of either irradiated fuel (IF) or reprocessing wastes (RW) arising from CANDU-PHW
(CANada Deuterium Uranium - Pressurized Heavy Water) reactors in Canada to the year 2015.
The IF concept considers disposal of 246 000 steel containers of irradiated fuel immobilized in
lead. The containers would be placed in a grid pattern in clay-backfilled rooms. The RW concept
considers the emplacement of 186 000 containers of reprocessing wastes immobilized in boro-
silicate glass. The containers would be emplaced in a grid of drill holes in the floor o f rooms
and the rooms backfilled. The rooms for both concepts would be arranged in panels at 1000 m
depth. The layout of the containers and rooms was based on a three-level-of-detail thermal rock
mechanics analysis programme which determined design parameters within a system of thermo
mechanical constraints. The resulting underground layouts, cost estimates and schedules are
given for each concept.
413
414 SCOTT and CHARLWOOD
1. GEOSCIENCE RESEARCH
1.1 Introduction
The Canadian geoscience research program for geological disposal of high-level radioactive wastes has evolved over the past six years from an initial request, early in 1973, by Atomic Energy of Canada Limited (AECL) to the Department of Energy,Mines and Resources (EMR) for geoscience advice on geological disposal. AECL's request arose from the requirements of their continuing responsibility for and research in nuclear waste management, which have been an integral part of the Canadian nuclear power program initiated in the early 1950's.
The initial request specifically included: identifyfactors for consideration in the concept of geological disposal, evaluate the United States' proposal for radioactive waste storage in salt, determine the potential of Canadian salt deposits for radioactive waste disposal, and examine the suitability of other geological formations in Canada for disposal of radioactive wastes. This latter task was considered to be the largest and one of particular significance as a complement to the extensive development work done on salt both in the United States and elsewhere.
Canadian geology, physiography and demography offer a wide range of choice of rock types as an alternative to salt. However, the choice of alternate rock types was also governed by the fact that the Province of Ontario was, and is anticipated to be, the major area for growth in nuclear power in Canada and that, with the resources available for geoscience research, all possible rock types could not be studied simultaneously. Therefore, the decision was taken to direct the main thrust ox geoscience research toward plutonic igneous rocks within the extensive area of the Precambrian Shield in Ontario and to continue the evaluation of sedimentary rock formations, but at a relatively lower level of research effort.
In 1975 the geoscience program, which was then conducted primarily within EMR, progressed from the prior assessment and evaluation of existing data and information to a preliminary field examination of igneous rock structures in Ontario, case history studies of engineering structures and mines in igneous rock with reference to the occurrence and distribution of structural discontinuities and groundwater flow, evaluation of exploration techniques and compilation of data on the thermal and mechanical properties of igneous rocks. Since 1975 the Canadian
PART I
IAEA-SM-243/68 415
geoscience program, as it pertains to igneous rock, has developed into an integrated component of the AECL.Fuel Cycle Waste Management Program involving federal and provincial government departments, industry and universities as shown in Figure 1.
1.2 Igneous Rock ProgramWithin the Canadian Fuel Cycle Waste Management Pro
gram, described by Hatcher et al. [1] and Boulton [2], verification of the concept of geological disposal constitutes the primary phase of the program followed by selection of a site for a demonstration vault, construction and operation of a demonstration vault and, ultimately, construction and operation of a commercial vault. Thus, the objectives of the geoscience program are to determine the suitability of igneous rock masses for geological disposal and to provide a rational basis for the selection of a rock type and geological setting for development of a demonstration vault.
For the purposes of concept verification and the establishment of a geoscience data base, attention is being directed toward four principal rock types: granite, anorthosite,syenite and gabbro in combination with structural settings ranging from highly to weakly fractured. Such rock types and structural conditions are relatively abundant throughout the Precambrian Shield of Ontario.
However, access to areas in Ontario for field research of igneous rock masses of interest for concept verification is subject to a prior approval process involving federal, provincial and municipal officials and to the implementation.of a public information program to ensure the best possible understanding by the public of the intent, purpose and scope of the field research program.
At present, and for the past several years, field research has been confined to the study of granitic rocks underlying AECL properties at Chalk River, Ontario, and Pinawa, Manitoba. Work at these localities will continue in 1979-80 and it is intended that the approval processes will enable field research to be undertaken during the current year on one or more additional igneous rock structures.
To attain the objectives of the geoscience program, four major complementary Activities have been established with each Activity containing a series of discipline-related Tasks as shown in Figure 2. Management of the program, which also provides for the co-ordination of work among the various program participants and for the evaluation of information and data arising from specific tasks, is related directly to the Activity/ Task structure.
Atomic Energy of Canada Limited, (2)Whitshell Nuclear Research Establishment
Department of Energy, Mines & Resources, /,>Science and Technology Sector
Environment Canada,Inland Waters Directorate
FIG. 1. Componen ts o f the Fuel Cycle Waste Managemen t Program.
416
SCOTT and
CHA
RLWO
OD
IÀEA-SM-243/168 417
TASKS TASKS TASKS
GEOLOGY-FIELD/OFFICE DRILLING, CORE LOGGING
BOREHOLE TELEVISION (TV) LOGGING
GLACIAL EROSION
ROCK PRO PERTIES
ACT IV ITY
TASKS
AIRBORNE SURVEYS-ELECTROMAGNETIC
AEROMAGNETIC (GRADIOMETER)
GRAVITY SURVEYS
KAGNETOTELLURIC SURVEYS
SURFACE ELECTRICAL SURVEYS
SURFACE SEISMIC
SEISMIC LATERAL BOREHOLEELECTRICAL AND STANOARD BOREHOLE SURVEYS
SEISMIC DOWNHOIE
GEOTHERMAL-LOGGING
REGIONAL AQUIFER STRAINBOREHOLE RADAR
BOREHOLE GAMMA-RAY
TOPOGRAPHIC MAPPING
SEISMIC RISK EVALUATION
BOREHOLE INSTRUMENT DEVELOPMENT
BOREHOLE PHYSICAL MEASUREMENTS
GEOCHEMICAL ANALYSES
(BOREHOLE, LAB)
HYDROGEOLOGICAL MODELLING
(CONTINUUM, FRAC. FLOW)
RADIONUCLIDE TRANSPORT
LABORATORY SUPPORT
THERMAL PROPERTIES
MAGNETIC PROPERTIES
MECHANICAL PROPERTIES
ELECTRICAL AND DYNAMIC
ELASTIC PROPERTIES
ROCK CRACK PROPERTIES
UNDERGROUND HEATER
BOREHOLE AND SHAFT SEALING
FIG.2. Geoscience program components.
1.3 Activity Objectives
Each of the four major Activities has a set of objectives related directly to verification of the concept of disposal of high-level radioactive waste in igneous rock that constitutes the principal objective of the geoscience program. Energy, Mines and Resources Canada, Science and Technology Sector, has responsibility for the Geological, Rock Properties and Geophysics Activities, and Fisheries and the Environment Canada, Inland Waters Directorate, has responsibility for the Hydrogeological Activity.
The Geological Activity has as its primary objectives the identification and evaluation of the physical, structural and petrological attributes of igneous rocks and associated geological materials; analysis of the regional structural setting and tectonic history of intrusive igneous rock structures; and evaluation of the type and magnitude of geological processes that may, in the long term, affect areas of the Precambrian Shield of interest for waste disposal.
418 SCOTT and CHARLWOOD
Primary objectives of the Rock Property Activity are: to develop and construct equipment, develop techniques, and apply methodologies arising therefrom for the testing of rocks under thermal and mechanical stress; to provide appropriate quantitative data on strength, deformation and thermo-physical properties of igneous rocks for vault design purposes; and to provide such additional rock property data as may be required by other Activities.
Within the Geophysics Activity the primary objectives are: to assist in the establishment of acceptance criteriarequired for validation of the concept of geological disposal; to provide appropriate exploration methodology that can be applied to potential vault sites that meet acceptance criteria; to develop interpretive techniques applicable to data arising from the non-standard use of airborne, surface and subsurface exploration geophysical methods; to provide regional interpretations of crustal stability; and to support and extend data and interpretations arising from studies within other Activities of the geoscience program.
Hydrogeological Activity objectives are: assessment and development of methodology for collection and analysis of field data relevant to groundwater flow and radionuclide transport from deep subsurface disposal zones; development of field and laboratory techniques to provide information on the origin, age, subsurface flow path and hydrochemical evolution of groundwaters in deep subsurface zones; and assessment of hydro- geological monitoring capabilities for the physical and hydrogeochemical evaluation of potential groundwater migration paths from subsurface zones of varying depths.
To meet the objectives of the various Activities, a series of Tasks have been developed as listed in Figure 2 and described by Scott [3]. Thus, the geoscience program comprises a spectrum of interrelated tasks designed to evaluate rock mass and rock substance attributes at both field and laboratory scales of investigation. Some of the Tasks within all of the Activities involve considerable research and development prior to their application as standard techniques within the program. Others are either direct applications or modifications of existing exploration and laboratory techniques having direct relevance to the program.
As may be expected,, analysis, interpretation and synthesis of the large amounts.of data arising from even the current level of program effort constitute a major task. Therefore, development of a user-oriented data management system is under Way. This system is expected to facilitate the effective
IAEA-SM-243/168 419
and efficient use of Task data and information for the purposes of: (1) providing a rational basis for selection of a combination of igneous rock type and structural setting that will provide a suitable geological environment for a vault, and (2) providing input to the pathways analysis which, in the larger scope of the Fuel Cycle Waste Management Program, will be used to examine interrelations between and effectiveness of multiple barriers within the complete waste disposal system.
1.4 Geoscience Research Highlights
A comprehensive summary covering the results of research from each of the Tasks within the program is beyond the scope of this paper. Accordingly, the following examples of research progress have been chosen to illustrate both the diversity of and innovation in geoscience research that has been generated by the issue of nuclear waste disposal.
1.4.1 Structural Geology
Structural discontinuities such as faults, fractures and shear zones are of primary significance both as potential pathways for radionuclide migration from a repository and in relation to the design and stability of underground openings. Analysis of aeriai photographs and other remotely sensed imagery in conjunction with surface geological mapping provides a means for relatively rapid identification of these structural features, particularly in areas of extensive outcrop. However, evaluation of these structural features is a complex task. Several techniques for fracture evaluation involving concepts of degree of fracturing, fracture length and degree of fracture interconnec- tivity have been developed by Brown [4]. A preliminary application of these, concepts to discontinuity data obtained from outcrops and from boreholes at the Chalk River, Ontario, research area shows a high degree of correlation (Fig. 3) between calculated values of fracture intersects and values of hydraulic conductivity obtained from borehole-packer tests. These discontinuity evaluation techniques are. still under development and require further statistical validation and application testing prior to demonstration of their operational usefulness.
1.4.2 Geophysical Surveying
Various airborne, surface and borehole geophysical surveying techniques have found ready application within the geoscience program as a complement to both regional and detailed structural geological studies. One such airborne geophysical method is the vertical gradiometer system that has been recently developed by the Geological Survey of Canada as described by Hood
420 SCOTT and CHARLWOOD
HYDRAULIC CONDUCTIVITY (c m /s )Ю'9 I 0 7 1СГ5 I0‘ 3
INTERSECTION VALUES
FIG.3. Comparison o f fracture intersect (I) values and hydraulic conductivity (from [4]).
FIG.4. Geological features interpreted from airborne gradiometer survey (from [5p.
IAEA-SM-243/168 421
et al. [5] and Kornik [6,7]. Although developed primarily as a technique for mineral exploration, the system, comprising two vertically separated tail-boom magnetometers with onboard digital data recording and analysis equipment, has a demonstrated capability for detecting short-wave-length, near-surface magnetic anomalies produced by both petrological and structural features. Delineation of structural features (Fig. 4), as interpreted from a measured vertical gradient map, is a useful complement to structural geological information obtained from other remotely sensed sources and from surface geological surveys.
1.4.3 Laboratory Studies of Core Specimens
The performance of a specific igneous rock type that may ultimately be chosen for the construction of a waste vault will be dependent not only on the structural features that may be observed or interpreted on the macroscale but also upon the pétrographie fabric of the rock substance and such discontinuities as may have developed on a microscale. Detailed laboratory examination of core specimens by microscopic and geophysical methods are thus important elements within the geoscience program.
As an example, visual examination of core specimens from the research area at the Whiteshell Nuclear Research Establishment, Manitoba, indicated the granitic rocks from this locality to be either massive or to have a weakly expressed foliation due to local concentrations of dark minerals. However, measurement of quartz optic axes in all specimens studied by Dence and Coles [8] showed a distinct tendency for the axes to be aligned so that they defined a plane and, for some specimens, to cluster around particular directions in that plane (Fig. 5). Quartz fa b r ic o r ie n ta t io n , such as encountered in the specimens from
Whiteshell, can have an influence on the accumulation of stress leading to possible failure under thermal or mechanical loads. Parallel laboratory measurement of magnetic susceptibility of Whiteshell core specimens further confirmed the anisotropic character of these rocks.
Rock substance discontinuities on the microscale, such as grain boundaries, pores and microfractures, have the potential for contributing to radionuclide migration by diffusion but, depending upon the geochemistry of the diffusion pathways, radionuclide retention may also occur. A study of these micro-discontinuities in terms of their occurrence and chemical properties is thus of importance in characterizing igneous rock masses.
To develop rapid measuring techniques to determine the permeability, porosity and pore structure of rocks, Katsube [9] has experimented with the laboratory application of electrical
422 SCOTT and CHARLWOOD
Increasing pole
concentration
FIG.5. Quartz crystal orientations, Drill Hole WN-I, Whiteshell Nuclear Research Establishment
(from Ref. [8\j.
WN* 1,2 FF/K
5r
о1Эо
2 -------------------------1----------1------------------------ 1----------1--------------------------1— ■— J
- I I -10 - 9 - 8
LOG OF PERMEABILITY (K) IN cm /s
FIG. 6. Formation factor versus permeability for granite and granodiorite samples from
boreholes WN-1, WN-2, Whiteshell Nuclear Research Establishment.
IAEA-SM-243/168 423
and acoustic measurements to core specimens. Preliminary results of these measurements on core specimens of granitic rocks from Whiteshell show a strong correlation (Fig. 6) between permeability and the electrical formation factor, the latter being a function of pore tortuosity, pore aperture and number of pores. Similarly, a strong correlation has been found between specimen porosity and the electrical formation factor. Thus, laboratory geophysical methods offer promise as a means for rapid evaluation of the desired rock substance parameters.
1.4.4 Hydrogeological Methods
Reliable methods for in-situ borehole measurements of fluid potential, hydraulic conductivity, groundwater velocity and determination of in-situ groundwater chemistry are essential to the geoscience program. To achieve the desired measurement reliability in a finely fractured igneous rock environment, adaptation and modification of available borehole equipment has been necessary and has constituted a significant part of the work being done by the Contaminant Hydrology Section, Inland Water Directorate, Fisheries and the Environment Canada.
Hydraulic tests have been conducted during drilling and in open boreholes at Chalk River and Whiteshell. These tests included shut-in pressure measurements and pressure-pulse transient tests during drilling, and constant pressure/constant flow- rate injection tests and pressure-pulse transient tests using straddle-packer equipment. Grisak [10] and Raven [11], in reporting on the development of testing equipment and procedures, have noted that rock mass hydraulic conductivities of the order of 10 7 to 10 11 cm/s may be measured by pressure-pulse transient testing, provided the test equipment is properly calibrated and test conditions carefully monitored. They also note that sensitive monitoring of surface and downhole .pressures and temperatures is essential for ensuring reliable measurement of hydraulic conductivities of 10 7 cm/s or less.
Sampling of groundwater from specific zones within boreholes has been facilitated by the use of multi-level ported casing designed and constructed by Westbay Instruments Ltd. of Vancouver. Use of this casing and a Martek Instruments Inc. borehole chemical probe has enabled the in-situ measurement of several geochemical parameters while continuously pumping to surface at low rates from an isolated zone of the borehole [10].
424 SCOTT and CHARLWOOD
Approximately 95 per cent of the Canadian landmass was glaciated during the Pleistocene Epoch; thus, the possible extent of erosion by future glaciations over areas that may be considered for waste disposal is a matter of significance for the geoscience program. Reported estimates of the amount of glacial erosion that has occurred on the Precambrian Shield vary from thousands of feet [12] to a few tens of feet [13].
In an attempt to resolve the reported disparity in extent of glacial erosion, Kaszycki and Shilts [14] have undertaken a study of glacial dispersion of distinctive sedimentary and felsic volcanic rocks of the Proterozoic Dubawnt Group within the District of Keewatin, Northwest Territories. Examination of approximately 800 till samples distributed over the dispersal zone, which extends 300 km from the Dubawnt outcrop belt to the shore of Hudson Bay, in combination with information on till thickness, has enabled a calculation of glacial erosion of the Dubawnt rocks of from 5.5 to 8.0 m.
While similar studies elsewhere on the Precambrian Shield may yield different results, it would appear that the previously reported lower estimate of glacial erosion may well be most realistic.
1.4.5 Glacial Erosion
PART I I
2. DESIGN CONCEPTS
2.1 Introduction
The current AECL design program considers the disposal of either immobilized irradiated fuel (IF) or solidified reprocessing waste (RW) [2]. Two different design concepts have been developed, one for each fuel cycle optiqn, and include the surface facilities [15,16]. Here, only the underground vaults are discussed. These waste emplacement and backfilling concepts were developed to illustrate the various alternatives available. They provide a basis for planning and preliminary safety studies and will be subject to complete review and possible modification in the future. The current concepts evolved from a range of preliminary designs proposed in a previous study [17].
IAEA-SM-243/168 425
The vaults have been designed to accommodate the projected irradiated fuel from the operation of CANDU-PHW reactors in Canada until the year 2015. This will require emplacement of either 246 000 IF or 186 000 RW containers using the preliminary immobilization and packaging concepts [15,16]. All wastes will have had a minimum of 10 years decay cooling before emplacement.
2.2 Preliminary Design Concepts
Preliminary layouts were developed prior to undertaking the detailed thermal rock mechanics analyses. The arrangement was based on an initial assessment of construction and operational considerations, and consisted of a modular system of panels about 400 m wide and 800 m long. The generalized layout is shown in Figure 7. Access to the vault would be provided via waste-handling and service shafts and haulage and ventilation drifts. A room-and-pillar (actually lane-and-pillar) type of excavation would be employed using conventional mining procedures.
Typical emplacement-room cross-sections are shown in Figures 8 and 9 for IF and RW emplacement, respectively. The IF would be immobilized in lead in stainless-steel containers,0.91 m in diameter, 1.15 m long. The RW would be immobilized in borosilicate glass in stainless-steel containers, 0.457 m in diameter, 3.235 m long. The heat generation rate per container would be 269 W at the time of emplacement for both IF and RW.
2.3 Thermal-Rock Mechanics Analysis
The thermal rock mechanics analyses established room and container layouts for IF or RW in terms of the specified constraints as summarized in Table I.
The thermal and mechanical constraints refer to three distinct geometric regions of the vault, which were analyzed separately :
The container near-field region including the container and the adjacent rock mass and backfill
The room-and-pillar region extending to several room diameters above and below the room
The far-field region extending from the ground surface to below the vault and beyond the edges of the vault.
SCOTT
and C
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WO
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IAEA-SM-243/168 427
TEMPORARY TOP OF BACKFILL
I OOm Í . OF CONTAINER
- ^ Г Г П CONTAINERm —r -LLL J 0.911
Jl.OOm j/II IÇZ3I1 K rv i
FIG.8. IF room cross-section.
FIG. 9. RW room cross-section.
The gross and panel thermal loadings1 are composites of several geometric variables, which allowed considerable latitude in selecting designs compatible with the constraints for granite or gabbro sites.
1 Gross thermal loading (GLT) = total initial heat generation of waste -r gross plan area of vault. Panel thermal loading (PTL) = initial heat generation in room net area of room and pillar.
428 SCOTT and CHARLWOOD
TABLE I. SPECIFIED THERMAL AND MECHANICAL CONSTRAINTS FOR
RW AND IF VAULTS
Geometric Region of Vault Thermomechanical Constraint RW Vault IF Vault
Container Container skin temperature 150°C 150 °Cnear-field (135 К rise) (135 К rise)
Container cavity stability open and stable hole
not applicable
Near-field rock mass stability
notsupported
not supported
Room and Backfill volume-average 100° С 100°CPillar temperature (85 К rise) (85 К rise)
Roof and rib conventional conventionalfailure/support rock-bolting
requirementsrock-boltingrequirements
Integrated average of strength-stress ratio in pillar
2 2
Far-field *Sustained long-term not 75°C to 125°Ctemperature applicable *(60 to 110 К rise)
*Rock mass stability negligible negligible irreversand deformation irreversible
deformation and perturbed fissure zone
<100 m deep
ible deformation and perturbed fissure zone <100 m deep
Ground surface movement comparable to long-term regional movements
comparable to long-term regional movements
* The constraints of sustained long-term temperature (75“C, etc.) and depth of perturbed fissure zone (100 m) evolved during the course of the investigations and have been treated as tentative values for initial study purposes only and are the subject of ongoing review.
The thermal-rock mechanics considerations included the in - s itu s tress, the geometry and strength of jo in ts , b last- fractured zones, fa ilu re charac te ris tics and nonlinear behaviour of the rock mass, and ind icated the requirements for support in the room-and-pillar region of the v au lt .
IAEA-SM-243/168 429
LEGEND
------------ G R A N IT E VAULT
------------G A B 8 R 0 VAULT
-2P T L = W / m ‘
FIG.10. IF vault design constraints program
SELECTED CONFIGURATION
О g r a n i t e v a u l t
G g a b b r o v a u l t
The allowable ranges of pitch and extraction ratio for an IF vault in granite or gabbro are summarized in Figure 10.For granite and gabbro, the preferred PTL-pitch-extraction ratio combination is governed by the thermal and/or thermal-mechanical performance in the far-field region in the long-term period, using the 75°C limit. A significant result of the IF analyses was that the average temperatures in the vault would be sustained for up to 15 000 years due to the uranium and plutonium content of the fuel. The basis for the selection of long-term criteria and restrictions to recognize this feature requires refinement.In particular, the significance of the depth of an allowable perturbed fissure zone requires research.
2 Pitch (P) = spacing of rows of containers along room. Extraction ratio (ER) = room
width -r room-and-pillar width.
430 SCOTT and CHARLWOOD
PITCH (m)
LEGEND
GRAN ITE VAULT
------------------G A B B R O VAULT
P T L = W / m 2
SELECTED CONFIGURATION
О G R A N IT E OR GABBRO
FIG.11. RW vault design constraints program.
The thermal-mechanical constraints and allowable ranges for pitch and extraction ratio for the RW vault concept are summarized in Figure 11. For both granite and gabbro, the preferred PTL-pitch-extraction ratio combinations are controlled by the minimum drill-hole spacing and the backfill temperature constraints. The temperatures should decrease to near ambient levels within 300 years. However, an increase in the backfill temperature constraint or reduction in the container skin temperature constraint could make this latter constraint significant, particularly for a vault in gabbro. Variations in effects of far- field constraints were not examined parametrically, but are not anticipated to be significant at the thermal loadings required by the other constraints. Room stability and pillar strength ratio criteria do not appear to be critical for any of the thermal loadings considered.
TABLE II. DESIGN PARAMETERS ADOPTED FOR RW AND IF VAULTS
IAEA-SM-243/168 431
Quantification of Criteria
Description of Criteria
Immobilized Fuel Vault (IF)
Reprocessing Waste Vault (RW)
LAYOUTTotal number of containers 246 475 186 300Depth of vault 1 000 m 1 000 mExtraction ratio (ER) 25 % 25%Room cross-section (width x height) 7.5 m x 6.15 m 7.5 m x 5.0 mContainer spacing across room 1.5 ra 1.5 mContainer spacing along room (pitch) 2 . 5 m (granite) 1.5 mShaft location outside vault 200 m 200 m
THERMAL LOADINGContainer power 269 W 269 WPanel thermal loading (PTL)
214.4 W/m (granite) 11.3 W/m^ (gabbro)
224 W/m (granite 24 W/m^ and gabbro)
Gross thermal loading (GTL)
14.2 W/m^ (granite) 11.0 W/m^ (gabbro)
23.6 W/m^ (graniteand gabbro)
MECHANICALBackfill at emplacement
Backfill to 3.15 m above floor
Complete backfill
Artificial support
Post-emplacementaccess
Conventional rock bolting in roof and walls of room Room access available until 2055, main drifts access until 2100
Conventional rock bolting in roof and walls of room Main drifts access until 2045
The design parameters adopted, which are consistent with the basic specifications and the results of the thermal rock mechanics analyses, are summarized in Table II.
2.4 Vault Layouts
The layout for the IF vault is shown in Figure 12.This vault would include 16 emplacement panels, each with 52 rooms containing up to 300 containers.
432 SCOTT and CHARLWOOD
FIG.12. IF vault layout (dimensions in m).
The RW vault layout is similar, except that the panel length is reduced to 430 m. This vault would include 16 panels, each with 24 rooms containing 500 containers.
The IF containers would be placed within the backfill to provide a geochemical barrier for long-term isolation. One metre of backfill would be placed above the container immediately after emplacement. The remainder of the room would not be backfilled until 20 years after completion of emplacement to allow access for in-room monitoring or possible retrieval.
The RW containers would be emplaced in holes drilled in the floor of the room. This concept simplifies the handling operations since the waste would be effectively shielded following emplacement and backfilling the drill hole. The pillar would be at least 7 m wide. This design concept assumes that the room would be backfilled completely after waste emplacement with a mixture of clay and crushed rock.
IAEA-SM-243/168 433
The width of the rooms (7.5 m for both IF and RW) is dictated by the container spacing (1.5 m) across the room. The heights of the rooms (6.15 m for IF and 5 m for RW) are governed by considerations of access and container handling, backfill depths (IF) and hole drilling (RW). The waste transport and panel drifts, as well as the main haulage drift, would be 4 m high by 5 m wide to allow for good tramming conditions for the ventilation trucks, and to provide an adequate cross-section for the ventilation airflow.
The following factors were also considered in determining the layout of the vault :
The layout employs a retreat system of mining and emplacement away from the heated areas and separation of mining and waste emplacement activities. This would allow uni-directional flow of ventilation air from the access shafts separately, through development and emplacement areas, to the exhaust shafts at the far end of the v ault.
Trackless diesel-powered equipment for conventional drill-and-blast excavation was considered for the most satisfactory performance and flexibility.
The size of the panel was determined by the optimum tramming distance for rock hauling and ventilation system requirements.
It was estimated that ambient rock temperatures would prevail beyond a distance of about 200 m from the storage rooms during the operational life of the vault. Therefore, the layout includes location of shafts at least 200 m from the nearest panel.
Initial development of all the major drifts would allow access for in-situ investigations around the entire area at the vault level. In addition, the retreat system would allow location of a demonstration panel at the exhaust end of the vault.
2.5 Development and Emplacement Schedules
The development and operation of the IF and RW vaults was considered in four phases,as shown in Figures 13 and,14, respectively.
PH A SE I
f Ï Í 1985 9 0 95 200 0 05
Ш
10 15 20 2 5 3 0 35 4 0 4 5
Ж
I Ф Щ f5 0 55 6 0 6 5 7 0 75 8 0 8 5 9 0 9 5 2Ю0 0 5
PR IM A R Y D E V E LO P M E N T A C T IV IT IE S
SHAFTU PC A ST S H A F T I U PC A ST S H A F T 2 DOW NCAST SER V IC E W A STESU RFA C E FAC IL IT IES
H A U LA G E W AYS:- M A INW ASTE H AULAGE I W ASTE H AU LA G E 2 ROCK R A IL H AU LAG E D R IFT B A C K F IL L ,R A IL H AULAGE DR IFTS
PA N EL D EVELO PM EN T 8 EM P LA C EM EN T ACTIVITIES
EXCAVATION ROOM PREPARATION W ASTE EM P LA C EM EN T BACKF ILLING ROOMS
M A IN T EN A N C E & DECOMMISSIONING ACTIV IT IES
MA INTENANCEBACKF ILL ING HAULAGE WAYS BACKF ILL ING SH A F T S RETR IEVAL OPTION
D EM O N ST R AT IO N VAULT
D EVELO PM ENT a OPERATION
K E Y TO M ILE ST O N E S (JAN I )
ф S T A R T C O N ST R U C T IO N '1984
(2) ST A R T D EM O N ST R AT IO N -1988
f?l START MAIN VAULT CONSTRUCTION-V 1994
^ ST A R T I F E M P L A C E M E N T -2 0 0 0
(§) C O M P LE T E IF EM PLA C EM EN T -2035
^ STA RT F IN A L ROOM BACKFILL-2055
(7) C O M PLETE F INALRO O M B A C K F ILL -V 2061
(Я) START VAULT B A CK F ILL IN G ANDV S E A L IN G -2 1 0 0
-------- tg) C O M P L E T IO N -2105
FIG. 13. IF vault developm ent schedule.
434 SC
OTT
and C
HA
RL
WO
OD
PHASE \____
U f §J y1985 9 0 95 2 0 0 0 0 5
12Ш
10 15 2 0 2 5 3 0 35 4 0 4 5 2 0 5 0
PR IM ARY D E V E LO P M E N T A C T IV IT IE S
SHAFT •- -UPCAST SH A FT I U P C A ST SH A FT 2 DOW NCAST SERV ICE W ASTESU RFA C E FAC IL IT IES
H A U LA G E WAYS MAINW ASTE HAULAGE I W A STE H AU LAG E 2 ROCK R A IL H AULAGE DRIFT BA C K F ILL ,RA IL HAULAGE DRIFTS
PA N EL D EVELO PM ENT & EM PLA C EM EN T ACTIVITIES
EXCAVATION ROOM PREPARATION W ASTE EM P LA C EM EN T BACKFILLING ROOMS
DECOMM ISSIONING A C T IV IT IES
BACKFILLING HAULAGE WAYS BACKF ILLING SH A F T S
DEMONSTRATION VAULT
D EVELO PM EN T a OPERATION
K E Y TO M ILE ST O N E S (JAN I )
ф S T A R T CO NSTRUCTIO N -1984
(2) ST A R T D EM O N ST R A T IO N -1988
[51 START MAIN VftULT CONSTRUCTION - V 1994
START RW E M P L A C E M E N T -2 0 0 0
l§) COMPLETE RW EM PLA C EM EN T -2044
C O M P L E T IO N - 2 0 5 0
FIG. 14. RW vault developm ent schedule.
436 SCOTT and CHARLWOOD
Phase I, access and demonstration, would require sinking of the exhaust shafts and development of rooms in which various experiments and equipment tests could be performed. If necessary, the central main drifts could be driven for exploration purposes.
Phase II, primary development, would follow satisfactory demonstrations. It would include sinking shafts in the main group and driving all haulageways, development of one panel to receive waste, and installation and commissioning of all container- handling and safety systems, backfilling equipment and underground services.
Phase III, continuing expansion and emplacement, would cover a period of 35 years for the IF vault. Following a 20-year retrieval option period, the rooms would receive the final backfill. Phase III for the RW vault would last for 44 years. The variations in these schedules reflect different container arrival rates and specified milestones for the alternative concepts.
Phase IV, the decommissioning of the underground facility, would involve backfilling all drifts and excavations, sealing all shafts and removal of surface facilities. The timing would vary for each concept.
2.6 Cost Estimates
The total capital and operating costs for construction, development and operation of the vaults are summarized in Tables III and IV for IF and RW vaults, respectively. These estimates do not include costs of demonstration tests and instrumentation. Room development and emplacement are classified as operating costs.
The total cost of the IF vault is estimated to be $1568 million, which is equivalent to $4.69/kg U or 0.10 mils/(kW-h) generated (1979 Canadian dollars).
The total cost of the RW vault is estimated to be $1174 million, which is equivalent to $4.35/kg U or 0.09 mils/(kW‘h) generated (1979 Canadian dollars).
Conclusions
Thermal-mechanical analyses indicate that satisfactory designs could be achieved using the preliminary design concept. Conventional mining practices could be used. Certain aspects
IAEA-SM-243/168 437
TABLE III. IF VAULT COST SUMMARY
(January 1979 Canadian Dollars)
Including allowances for additional items, administration and contingencies
Phase Timing Capital ($ x 1000)
Operating ($ x 1000)
Total ($ x 1000)
I 1983-1992 27 673 - 27 673
II 1993-1999 138 129 - 138 129
III 2000-2034 67 811 802 132 869 943
IV 2035-2105 20 806 511 784 532 590
TOTAL 254 419 1 313 916 1 568 335
TABLE IV. RW VAULT COST SUMMARY
(January 1979 Canadian Dollars)
Including allowances for additional items, administration and contingencies
Phase Timing Capital ($ x 1000)
Operating ($ x 1000)
Total ($ x 1000)
I 1983-1992 25 760 - 25 760
II 1993-1999 132 844 - 132 844
III 2000-2044 75 843 804 478 880 321
IV 2045-2049 2 828 132 790 135 618
TOTAL 237 275 937 268 1 174 543
438 SCOTT and CHARLWOOD
will benefit from development, e.g., hole-drilling equipment and backfill-design and -placing systems, but no insurmountable problems have been identified to date. The rock stabilization requirements during construction and emplacement are quite standard, and the thermal perturbations to room stability are minimal.
Considerable flexibility exists in the design concepts to accommodate waste packaging, backfilling and sealing innovations as they are developed, as well as changes of schedule or quantity of wastes. The temperature limits could be adjusted up or down by simple layout changes.
ACKNOWLEDGEMENTS
The author of Part I is grateful for the helpful comments of his colleagues: G.E. Larocque, J.G. Tanner andB.V. Sanford in Energy, Mines and Resources, and G.E. Grisak in Fisheries and the Environment.
The contents of Part II of this paper are based on the results of studies performed by Acres Consulting Services Limited in association with RE/SPEC Inc. of Rapid City, S.D., for Atomic Energy of Canada Limited. The AECL technical coordinator wasH.Y. Tammemagi. The contributions of the author's co-workers, including A.S. Burgess, P-0. Sandstrom and M.A. Mahtab of Acres and P.F. Gnirk and J.L. Ratigan of RE/SPEC, are gratefully acknowledged. The views expressed are those of the author and do not necessarily represent those of Atomic Energy of Canada Limited.
REFERENCES
[1] HATCHER, S.R., MAYMAN, S.A., TOMLINSON, М., "The Canadian programme for the development of deep underground disposal for nuclear power waste", these proceedings, SM-243/167.
[2] BOULTON, J. (ed.), Management of Radioactive Fuel Wastes:The Canadian Disposal Program, Atomic Energy of Canada Limited Report, AECL-6314 (1978).
[3] SCOTT, J.S., EMR Program for Geological Disposal of High-Level Radioactive Wastes, Disposal of High-Level Radioactive Waste: The Canadian Geoscience Program, Geological Surveyof Canada, Paper 79-10 (1979).
[4] BROWN, P.A., Fracture Studies at Chalk River, Ontario, in progress.
IAEA-SM-243/168 439
[5] HOOD, P.J., SAWATZKY, P., KORNIK, L.J., McGRATH, P.H., Aeromagnetic Gradiometer Survey, Atomic Energy of Canada Limited Technical Record, TR-11 (1979).
[6] KORNIK, L.J., Vertical gradiometer data of use in uranium search, Northern Miner (30 Nov. 1978).
[7] KORNIK, L.J., Vertical gradiometer interpretation, Lac Du Bonnet Pluton, in progress.
[8] DENCE, M.R., COLES, R.L., Rock crack properties, 1978/79 summary report in progress.
[9] KATSUBE, T.J., Laboratory studies, 1978/79 summary report in progress.
[10] GRISAK, G.E., Hydrogeologic research activities, 1978/79 summary report in progress.
[11] RAVEN, K.G., Studies in fracture hydrology at Chalk River Nuclear Laboratories, 1977/78/79, in progress.
[12] WHITE, W.A., Deep erosion by continental ice sheets, Geol. Soc. Amer., Bull. 83 (.1972) 1037.
[13] FLINT, R.F., Glacial and Quaternary Geology, John Wiley and Sons Inc., New York (1971).
[14] KASZYCKI, C.A., SHILTS, W.W., Average Depth of Glacial Erosion, Canadian Shield, Current Research, Geological Survey of Canada, Paper 79-1B (1979).
[15] ACRES Consulting Services Ltd. and Associates, and WNREDesign and Project Engineering Branch and Associates, A Disposal Centre for Irradiated Nuclear Fuel: ConceptualDesign Study, Atomic Energy of Canada Limited Report, AECL-6415 (1979).
[16] ACRES Consulting Services Ltd. and Associates, and WNREDesign and Project Engineering Branch and Associates, A Disposal Centre for Immobilized Nuclear Waste: ConceptualDesign Study, Atomic Energy of Canada Limited Report, AECL-6416 (1979).
[17] ACRES Consulting Services Ltd. and Associates, Radioactive Waste Repository Study, Parts I, II and III, Atomic Energy of Canada Limited Report, AECL-6188-1, 2, 3 (1978).
DISCUSSION
K. KÜHN : Dr. Charlwood mentioned in his presentation that the rooms
would be backfilled by “a geochemical bulk of material” . In view of the great
importance of backfdling, especially in the case of the disposal of spent CANDU
fuel elements for a very long period, I should like to ask what types of backfilling
material are being investigated. I wonder if anyone is considering a material other
than bentonite.
440 SCOTT and CHARLWOOD
R.G. CHARLWOOD: Our design studies were principally concerned with
construction and operational aspects of backfilling, and we used an arbitrary design
(similar to the KBS design) of bentonite, clay and coarse materials. CANMET
and AECL-WNRE are investigating these questions and are looking at a wide range
of materials; their findings are to be published soon.
D. A. GRAY : Dr. Scott indicated that the erosion risk was minimal in many
areas of the Canadian shield. Yet the design depth given by Dr. Charlwood was
1000 m. What is the main reason for considering such a depth, which is significantly
greater than that in many proposals?
R.G. CHARLWOOD: We are considering a conservative mining case in the
sense that 1000 m is beginning to approach the lower limit for acceptable simple
mining procedures in the in situ stress fields expected. Having demonstrated the
feasibility of construction at 1000 m, we would confidently expect that construc
tion at 500 m was also achievable.
F.L.H. LAUDE: Do you intend to use air-cooling by forced or natural con
vection when the repository is full? If so, for how long?
R.G. CHARLWOOD: No. Our design concept provides for sufficient
dispersion of the waste to allow heat conduction through rock to maintain
temperatures within required limits. Moreover, we wished to use “passive”
systems only; these do not require long-term operation and maintenance as
do air-cooling systems.
J. HAMSTRA: You have studied both spent fuel and reprocessing waste
disposal for CANDU fuel, and mention a heat production period of 20 000 years
in the case of the spent fuel option. Do you have a figure for the difference in
total integrated heat load between the two options?
R.G. CHARLWOOD: These figures are included in our Technical Records
which are to be published by AECL.
J. HAMSTRA: Do you also have a figure which indicates how this difference
in heat load would affect repository size for equivalent amounts of spent fuel and
reprocessing waste?
R.G. CHARLWOOD: The size is directly related to the gross and panel
thermal loadings, namely 14 W/m2 for IF and 24 W/m2 for RW (see Table II).
F.S. FEATES: Could you comment on the relative safety of the disposal
of spent fuel as compared with that of reprocessed waste? What would be the
effect of surface storage for 50 years (i.e. until the fission product heating has
passed its peak)?
R.G. CHARLWOOD: The next phase of the AECL programme will involve
quantitative comparisons of concepts based on safety analyses. To date we have
only made qualitative comparisons and are keeping our options open. However,
certain key factors have emerged:
IAEA-SM-243/168 441
(1) In terms of operator doses, handling waste in boreholes appears to be simpler
than room emplacement of small-diameter cylinders;
(2) Isolation by emplacement in drillholes may also have advantages in terms of
hydrogeological separation from the room and the associated disturbance
of its rock pore;
(3) An investigation should be made of other in-room concepts, possibly involving
the use of fabricated “drill” sleeves to improve shielding.
Thermal studies indicate that for RW the controlling factor is the mechanical
spacing of 1.5 m. Cooling would be of no significant rock mechanics benefit here.
For IF some benefit of closer spacing may be available if cooled fuel is
emplaced. However, I question the economic benefit of cooling in a separate
facility.
Y. INOUE: Who will construct and operate the repository in Canada? Who
will pay for it, and who will be responsible for environmental safety?
J.S. SCOTT: It is expected that the Federal Government will have overall
responsibility for an operational repository. So far no decision has been taken
concerning the institutional arrangements for the actual construction and operation
of a repository.
The Atomic Energy Control Board (AECB) will have responsibility for
licensing the construction and operation of a repository. Thus the owner of
the repository must satisfy the AECB requirements concerning safety. Environ
mental assessments will also be prepared by the repository owner to meet the
requirements of environmental departments of the appropriate levels of Government
jurisdiction.
Y. INOUE: Will you please explain the pressure pulse transient tests used
to measure hydraulic conductivity?
J.S. SCOTT: Pressure pulse transient testing is a borehole technique in
which a section of the borehole is isolated between packers and a mechanical
or hydraulic load is imposed upon the water-bearing zone between the packers.
Through the use of transducers or other pressure-sensing devices and recording
equipment a record of the pressure build-up and decay is obtained. Permeability
values are obtained from analyses of the pressure decay curve.
IAEA-SM-243/163
THE SWEDISH GEOLOGICAL PROGRAMME FOR THE DISPOSAL OF HIGH-LEVEL WASTE
U. THOREGREN, K. AHLBOM, G. GIDLUND,C.E. KLOCKARS, K.A. MAGNUSSON,S. SCHERMAN Geological Survey of Sweden,Uppsala,Sweden
Abstract
THE SWEDISH GEOLOGICAL PROGRAMME FOR THE DISPOSAL OF HIGH-LEVEL
WASTE.
During the last two years the Geological Survey of Sweden (SGU) has fulfilled a very
intensive geological programme on behalf of the Nuclear Fuel Safety Project (KBS). This work
is now completed and SGU’s geological programme for radioactive waste disposal in crystalline
rock is now in a second phase. This includes site selection studies in both southern and
northern Sweden, comprising geological, geophysical and hydrological studies of selected
sites. In addition to this work a special research area will be used for detailed investigations
concerning methods for site confirmation. This area is situated at Finnsjôn near Forsmark.
1. INTRODUCTION
During the last two years, the Geological Survey of Sweden (SGU) has
fulfilled a very intensive geological programme on behalf of the Nuclear Fuel
Safety Project (KBS).
This project was formed by the nuclear power utilities in Sweden and SGU
was given the task of meeting the requirements of the ‘Stipulation Law’. This is
a law “concerning special permission for charging nuclear reactors with fuel”
which was enacted by the Swedish Parliament in April 1977.
The principal geological investigations have been performed at three sites,
Karlshamn, Oskarshamn and Forsmark. In addition geological, rock mechanical
and tracer tests have been carried out in the abandoned Stripa Mine and at
Studsvik. The locations of the areas studied are shown in Fig. 1.
The studied areas contain our most common types of rocks, namely granite,
gneiss and gneiss-granite. These areas are representative for the geological con
ditions of large parts of southeast Sweden [1-4].
At Karlshamn, Oskarshamn and Forsmark, 17 boreholes with core recovery
were drilled by SGU down to a depth of about 500 m. Altogether about 10 000 m
of recovered core have been examined [1, 2].
443
444 THOREGREN et al.
FIG.l. Locations of study areas for KBS (black circles) and for PRA V (open circles).
The hydraulic conductivity of rock mass along boreholes was determined
from water pressure measurements. Some 12 000 measurements have been
carried out in the boreholes [5, 6]. In addition some of the boreholes have been
investigated by different types of geophysical logging methods [5].
To determine age and chemical composition of groundwater in the investi
gated areas, groundwater samples have been collected from different depths in the
boreholes. The age determination was carried out with the carbon-14 method
[7].Field tests have been conducted by SGU at Studsvik to study the retardation
of the radioactive elements as they are transported with the groundwater through
fissured rock [8].
SGU has now concluded the programme for which it was commissioned by
KBS. In the immediate future the aim of the Survey is to continue the site
selection studies in both southern and northern Sweden.
In connection to these studies a special research area will be used at
Finnsjôn near Forsmark (Fig.l). In this area the whole hydrological cycle will be
studied. Boreholes will be drilled, in which new geophysical and hydrogeological
instruments are measuring methods will be tested.
Both at Studsvik and in the Stripa Mine tracer tests will be performed to
obtain a better understanding of the migration of radioactive nuclides in
crystalline rocks.
Some details of this geological programme concerning the disposal of
radioactive waste will now be presented .
IAEA-SM-243/163 445
During the summer of 1978 the Swedish Geological Survey (SGU),
commissioned by the National Council for Radioactive Waste (PRAV), carried
out geological investigations in order to find places suitable for a future repository
for high-level radioactive waste.
The first stage of this investigation concentrated on southern Sweden.
The work was carried out in two main phases:
( 1 ) General investigations comprising studies of geologic literature and aerial
photo interpretation.
(2) Fieldwork to establish geologic and tectonic maps of choosen areas.
Genera] literature studies concerning radioactive waste in hard crystalline
rock provided guidance.
With experience from underground construction and tunnelling in Swedish
crystalline bedrock the following aspects were considered important for obtaining
high quality in any repository planned:
low rock-surface topography;
low frequency of joints and faults;
high degree of rock uniformity;
low water-content in the rock-mass (indicated by drilled wells).
In order to establish a broader selection of rock types, formations which
could only fulfil some of these criteria were also taken into consideration.
The following bedrock formations have been choosen for future study (the
figures in brackets indicate the position of these in Fig.l):
veined gneisses on the south west coast (1);
strongly metamorphosed gneisses and granites of south west Sweden (2);
veined gneisses in the south east of Sweden (3);
coastal gneisses on the southern coast of Sweden (4);
granitic formations of different ages and tectonic setting in central
southern Sweden (5) and on the west coast (6 and 7).
This first step in the investigations of the suitability of Swedish rock
formations for a radiation waste repository was made as a pilot study. Some
characteristics from different formations are as follows.
( 1 ) The granites show quite distinctive and pervading fracturing with at
least three sets of intersecting fracture zones present. That means a
good hydraulic connection between fracture sets. In younger granites
(see Nos. 6 and 7 on the map) the fracture sets are usually water
bearing, as shown by drilled wells. At greater depths large sections of
practically dry rock were often found between such fracture zones.
2. SITE SELECTION
446 THOREGREN et al.
(2) Migmatites and veined gneisses have proved favourable in underground
construction. The structures of these rocks reflect plastic folding and
migmatitic veining and therefore counteract regular and pervading
fracture sets. This leads to a decrease in hydraulic conductivity.
The programme for future work is as follows:
Further prospecting for favourable bedrock formations. This work is
intended to cover additional rock types and the programme will also be
extended to the northern parts of Sweden.
Selection of suitable sites for further geological, geophysical and hydro-
logical investigations.
Geophysical investigations applied at sites selected for further geological
investigations. In time these will be followed by exploratory drilling.
Interpretation of geological and geophysical data in relation to the
constructional and hydrogeological properties of the bedrock.
Survey of international developments in the entire field through literature
and contacts with other groups working on similar problems.
Airborne measurements (altitude of flight 30 m), covering a more extensive
area around the selected site, are used in order to get a regional picture of
structures and geological setting. The airborne measurements of the
Geological Survey of Sweden use the following methods:
Magnetic
VLF
Radiometric.
The VLF measurements are used for detecting major clay-filled or water
bearing fissure zones. The magnetic measurements give valuable information
about structures and dislocations. Radiometric measurements are used for
mapping of rocks and structures.
Surface geophysical investigations are used to obtain more detailed informa
tion about the physical conditions at the site and its surroundings. The methods
are chosen according to prevailing physical conditions of the site. In most cases
the following sets of methods are used:
resistivity and induced polarization (IP);
slingram (moving source — receiver; EM-method);
VLF;
seismic refraction.
Resistivity, IP and slingram are measured in a grid net from which maps are
constructed. In areas of interest VLF and seismic profiles augment the above-
mentioned methods. Resistivity and slingram methods indicate water-bearing or
IAEA-SM-243/163 447
clay-filled fracture zones. The VLF method is used to discriminate between
major and minor zones. Seismic measurements are used for rock-quality
determination.
3. METHODS FOR SITE CONFIRMATION
3.1. Hydrogeological test-site in Finnsjón
At Finnsjón, close to Forsmark, see Fig.l, a long-term research area will be
established.
The overall aim with this research area is to evaluate what factors are of
importance for the selection of a site for radioactive waste disposal and to
determine how to confirm the suitability of that site. Methods for this purpose
and for investigations over greater volumes of bedrock will be developed in this
project.
The Finnsjón area, which consists of a 25 km2 catchment area, will be
investigated very thoroughly. Some of the work has already started, for example,
soil mapping [4], bedrock mapping and some hydrological mapping. More
detailed bedrock mapping and fracture mapping will be performed this summer.
Studies of the water balance and maps of inflow and out flow areas will be
established.
The groundwater flow will be followed from infiltration to the outflow areas
through studies in shallow and deep boreholes.
For the moment there are seven deep boreholes (500 m) and eight shallow
boreholes (70 m) in this area. A couple of additional boreholes will be drilled to
complete some interesting profiles for the groundwater flow in the area.
The chemical composition and age of the groundwater will be used to
investigate the history of the water flow.
One of the items of interest is the development of methods and instruments
for crosshole techniques. The aim is to find techniques that can determine rock
conditions at greater distances from the boreholes. This allows a three-dimensional
evaluation of the rock properties.
3.2. Methods and instrumentation
In previous work carried out by SGU on storage of radioactive waste, several
methods have been applied and the evaluation of these methods is in progress.
SGU has chosen to úse as many methods as possible in order to select a suitable
set for future work in this field.
It is important to apply a set of methods which are favourable as regards
the actual physical and hydrological properties of different sites. In order to
448 THOREGREN et al.
interpret the data correctly, different methods which complement each other are
needed.
3.2.1. Geophysical methods and instrumentation
The aim of the geophysical investigation is to determine different rock
properties using a combination of methods. Detailed information on the rock
properties close to the borehole can be obtained by the following methods:
TV-inspection;
point-resistance (detection of water-bearing zones);
differential point resistance (open fractures);
natural gamma (rock properties).
Information on rock properties within a couple of metres from the borehole
can be obtained by the following methods:
resistivity (detection of water-bearing zones);
IP (mineralization).
Methods with greater depth penetration are:
slingram (conductive zones);
VLF (conductive zones).
The condition of the borehole water is studied by:
resistivity of borehole fluid (salinity);
pH;
Eh.
Furthermore, the following methods have been applied: the borehole
deviation, temperature and self-potential (SP).
The above-mentioned methods can in future be complemented with other
methods, for instance, acoustic, neutron and density.
In the future, cross borehole techniques such as the following will be used:
borehole seismic studies;
electromagnetic methods (high frequency tools);
electrical methods (mise-à-la-masse).
A promising tool is the borehole radar technique which gives information
about the three-dimensional extension of the major fracture zones.
New ion selective electrodes for in-situ determination of the water quality
are being developed in co-operation with the Royal University of Technology,
Stockholm. These electrodes can be used in connection with a groundwater pump,
which is described in the following test.
IAEA-SM-243/163 449
SGU has mainly used the hydrological method for estimating hydraulic
conductivity. To date SGU has measured a total length of about 12 000 metres
of deep boreholes, drilled in crystalline rock. These investigations have been
carried out in very competent rock formations. Therefore equipment for the
determination of very low flow and giving very accurate pressure readings had to
be developed. The equipment used allows measurements of hydraulic conductivity
as low as 10~n m/s (3 metre sections). Three different pressures have been used
in these measurements (0.2, 0.4 and 0.6 MPa). Both single and double packer
configurations have been used.
Another method used to investigate the suitability of a site is to determine
the age and chemical composition of the groundwater. The sampling is carried
out with a pump placed between the packers. Thus a fracture zone of interest
can be sealed off for water sampling. The methods used for age determination are
carbon-14 and tritium analyses. Development of methods and instruments is in
progress.
The packer equipment has been rebuilt in order to eliminate leakage around
the packers. To focus the flow in a horizontal direction equipment with multiple
packer configuration is under development. This is also needed to obtain proper
results for future piezometric measurements.
In future investigations the pulse-pressure technique will be applied. New
methods for age determination of groundwater are being studied and a set of
different methods for age determination will be applied. In future work this will
make it possible to compare results in order to eliminate the errors which can
occur in different environments.
To improve the quality of water analyses a pressure tank will be used on
the surface for collecting the gas-phase of the depressurized water. The gas-phase
will be analysed in a gas-chromatograph.
3.2.2. Hydrological methods and instrumentation
3.3. Tracer test
3.3.1. Hydraulic conductivity o f fissures
By regarding the rock as homogeneous, with average hydraulic conductivity
including fissures and more or less impermeable rock-mass, the quantity of the
water flow can be described. The real velocity of the water is however dependent
on the hydraulic conductivity of the fissures. By using tracers it is possible to
measure the velocity of the groundwater and in some case also to estimate the
kinematic porosity. The kinematic porosity is commonly used to describe the
relation between hydraulic conductivity of the rock-mass and of the fissures.
450 THOREGREN et al.
B2
FIG.2. Cutaway view of the in situ experimental area.
The theory for determining the groundwater velocity by using artificial
tracers is quite simple. Connection between two or more boreholes through a
fissure must be established. The tracers are injected in a sealed part of the fissure
in one borehole. By pumping water from another borehole it is possible to
control the flow-pattern and thus measure the rate of the tracer transport along
the fissure.
Tracer tests were performed in two localities. At Studsvik the bedrock was
a granitic gneiss, in which a horizontal fissure was studied [8]. A cutaway view
of the in situ experimental area is shown in Fig.2. At Finssjôn the bedrock was
granodiorite and the fissure subhorizontal. In the first locality different radio
nuclides were used as tracers. 82 Br was assumed to behave like water and
was used as a reference-tracer. In Finnsjôn Rhodamine WT was used as tracer.
IAEA-SM-243/163
TABLE I. BOREHOLE DATA: STUDSVIK (82Br)
451
Flow path B8 — B2 B8 — B7
Distance between boreholes 51 m 22 m
Transit time 36 h lO h
ks 7.6 X 10"6 m/s
kf 6.5 X 10-3 m/s 6.1 X 10-3 m/s
Kinematic porosity 1.2 X 10 '3 1.2 X 1(T3
TABLE II. BOREHOLE DATA: FINNSJÓN (Rd WT)
Distance between boreholes 35m
Transit time 21.5 h
ks 2.2 X 1СГ6 m/s
kf 1.4 X I O' 3 m/s
Kinematic porosity 1.6 X 10-3
Time (hours after injection)
FIG. 3. Concentration curves of S7Br and isSr for the flow path В 8 — В 2.
452 THOREGREN et al.
In Tables I and II data are given of the boreholes, measured transit-time and
distances for the localities investigated. The breakthrough-curves for one of the
tests are shown in Fig.3. In Tables I and II values are given for the mean
hydraulic conductivity both of the fissures kf and of the 2 m section which seals
off the fracture zone ks. The former was calculated from the tracer test. In the
table the ratio obtained for the two types of hydraulic conductivities is given as
kinematic porosity.
3.3.2. Radionuclide migration tests in fractured rock
Radionuclides representing long-lived fission products of the elements
selenium, technetium, tin, cesium, iodine, neodymium and strontium were
studied in field experiments at Studsvik. The nuclides were injected into a
fracture zone intersecting one of the test holes at a depth of 72 metres below the
ground surface. Concentration-versus-time curves of activities arriving at the
pumping borehole were measured. The mean transit time and the dispersion
coefficient for water were calculated to be 57 h and 8.3 m2/h, respectively.
The radioactivity measurement techniques included gamma-spectrometric
borehole logging, continuous activity measurements of pumped water and
water sampling followed by laboratory analysis. Different combinations of
radionuclides were injected simultaneously and selectively registered by gamma-
ray spectrometry using a Ge(Li) detector in the field as well as in the laboratory.
Technetium and iodine travelled as anions with the same velocity as that
of water. Strontium was retarded by a factor of about 6. Neither cesium nor
neodymium could be detected.
4. MODEL STUDIES
In two of the investigated areas (Finnsjôn and Sternô) mathematical models
have been used in the study of groundwater movements.
The Sterno area has been analysed by means of two different two-
dimensional radial-symmetrical models. In both models flow lines, flow times and
gradients of the groundwater in a number of profiles have been calculated, with
topography, permeability and porosity of the rock mass as input data. The
effects of large fracture zones have been taken into account in the calculations in
one of the models.
In the Finnsjôn area a three-dimensional numerical model in accordance
with the fínite-element method has been used. This model gives results which
are more realistic than the two-dimensional models but is also more complex.
IAEA-SM-243/163 453
REFERENCES
With the exception of Ref.[8] these are in Swedish.
[1] SCHERMAN, S., Fôrarbeten for platsval, berggrundsundersokningar. KBS Teknisk
Rapport 60, Stockholm (1978).
[2] OLKIEWICZ, A., SCHERMAN, S., KORNFÀLT, К-A., Kompletterande berggrunds-
undersôkningar inom Finnsjô- och Karlshamnsomrâdena. KBS Teknisk Rapport 79—05,
Stockholm (1979).
[3] OLKIEWICZ, A., HANSSON, К., ЛЕМЕХ К-E., GIDLUND, G„ Geologisk och hydro-
geologisk grunddokumentation av Stripa forsoksstation. KBS Teknisk Rapport 63,
Stockholm (1978).
[4] ALMÉN, К-E, EKMAN, L, OLKIEWICZ, A., Fôrsôksomrâdet vid Finnsjôn. Beskrivning
till berggrunds-och jordartskartor. KBS Teknisk Rapport 79—02, Stockholm (1979).
[5] HULT, A., GIDLUND, G., THOREGREN, U., MAGNUSSON, K-Â., DURAN, O.,
Permeabilitetsbestamningar. Geofysisk borrhâlsmatning. KBS Teknisk Rapport 61,
Stockholm (1978).
[6] GIDLUND, G., HANSSON, K., THOREGREN, U., Kompletterande permeabilitets-
matningar i Karlshamnsomrâdet. KBS Teknisk Rapport 79—06, Stockhom (1979).
[7] GIDLUND, G., Analyser och âldersbestâmningar av grundvatten pâ stora djup. KBS
Teknisk Rapport 62, Stockholm (1978).
[8] LANDSTRÓM, O., KLOCKARS, К-E, HOLMBERG, К-E., WESTERBERG, S., In Situ
Experiments on Nuclide Migration in Fractured Crystalline Rocks. KBS Teknisk
Rapport 110, Stockholm (1978).
DISCUSSION
P.A. WITHERSPOON {Chairman)-. What tracers did you use in your
work:
U. THOREGREN: At Studsvik, apart from 82Br, the other isotopic tracers
studied were 24Na, 75Se, 85Sr, "Tcm, 113Sn, 131I, 131Cs and 147Nd. At Finnsjôn
an inactive tracer (Rhodamine WT) was used-.
C. DAVISON: You mentioned that you have carried out hydrogeologic
tests in boreholes to depths of 600 m or more. Have your tests indicated that
there is active groundwater flow occurring at these depths?
U. THOREGREN: In several boreholes we have measured hydraulic
conductivities of as much as 10-6 m/s at depths of 500—600 m. TV
investigations of the boreholes at these depths have also revealed open fractures.
The groundwater from these depths varies between 5000 and 13 000 years in age.
All these results indicate an active groundwater flow at the depths mentioned.
G. STOTT : May I inquire what is the approximate cost of the impressive
geological programme you described, if possible in terms of its phase 1 and
phase 2?
454 THOREGREN et al.
U. THOREGREN: The intensive geological programme being carried out
on behalf of the KBS and the National Council for Radioactive Waste Manage
ment costs about 18 million Swedish crowns or US $4 million (phase 1). The
cost of phase 2 (site selection and site confirmation) is about 15 million
Swedish crowns of US $3.5 million per annum.
L.J. ANDERSEN: In using the TV inspection technique, how do you
quantify the results?
U. THOREGREN: The fractures are classified according to their estimated
aperture and divided into three groups: I. < 1mm, II. 1 — 5 mm, III. >5 mm.
The detailed data will be found in KBS Technical Report No. 61 (see Ref.[5]).
A. AZIZ: With reference to the borehole radar technique determining
three-dimensional extension of fissure zones, how, and to what extent, can this
method be used in the geological formations?
U. THOREGREN: The radar equipment for borehole measurement has
recently been developed in Canada. It may provide valuable information about
the fissure zones close to the boreholes. In high-resistivity rocks it has an
estimated range of about 60 m. In rocks of low resistivity the range of penetration
will decrease markedly.
F. GERA: Have you considered the use of the acoustic televiewer? It is
particularly effective for mapping fractures intersected by boreholes.
U. THOREGREN: We may use this device in our future work. We consider
the method to be very useful in many respects.
IAEA-SM-243/29
REVIEW OF GEOLOGICAL CRITERIA AND
SITE SELECTION FOR HIGH-LEVEL
RADIOACTIVE WASTE REPOSITORIES
IN THE UNITED KINGDOM
J.D. MATHER, D.A. GRAY,
P.B. GREENWOOD
Institute of Geological Sciences,
London,
United Kingdom
Abstract
REVIEW OF GEOLOGICAL CRITERIA AND SITE SELECTION FOR HIGH-LEVEL
RADIOACTIVE WASTE REPOSITORIES IN THE UNITED KINGDOM.
Systematic geological studies directed towards the disposal of highly radioactive heat-
emitting wastes have been pursued in the UK since 1975. Eight geological criteria have been
proposed for the selection of potentially suitable areas in the UK and these are reviewed in
the context of other international contributions to the debate on the feasibility of geological
disposal. Application of these criteria results in the recognition of over 100 different areas
containing potentially suitable host rocks. At the present time exploratory drilling has been
initiated at one site within a crystalline rock area. It is concluded that the principal require
ment is site-specific field data on which to base decisions on the most appropriate geological
environment for a repository.
INTRODUCTION
The problem of the disposal of waste produced from nuclear laborato r ies and plants was appreciated at an e a r ly stage in the United Kingdom (1) and the Geological Survey - now the In s t i tu te of Geological Sciences ( IGS) was f i r s t involved in ad hoc studies of potent ia l s i te s in 1951. However, these e a r ly a c t i v i t i e s were p r in c ip a l l y concerned with the disposal of waste of very low a c t i v i t y and only since 1975 have systematic geological studies been d irected towards the disposal of h igh ly rad ioac t ive heat-emitting wastes. At that time the United Kingdom Atomic Energy Authority (UKAEA) commissioned IGS to define the geological c r i t e r i a , sp e c i f i c to the United Kingdom, to be used in the se lec t ion of areas containing geological formations p o te n t ia l ly su i tab le for the disposal of h igh ly rad ioac t ive so l id wastes ( 2)and to se le c t areas which met those c r i t e r i a . A more deta i led review of areas containing geological formations p o te n t ia l ly su i tab le for the disposal o f h igh-level and/or a -emitting rad ioac t ive wastes was compiled as the UK component of a Catalogue being prepared by the national geological organisations of the Member States of the European Community.
455
456 MATHER et al.
The pre lim inary desk, studies have ted to a major research e f f o r t to examine the f e a s i b i l i t y of geological disposal of h igh-level wastes in the geological environment of the UK - research ca l led for by the Royal Commission on Environmental Po l lu t io n (3 ) . E f fo r t has so fa r been concentrated in the c r y s t a l l i n e rocks although no decision has been made on the most appropriate host-rock and p a ra l le l studies are now being in i t ia te d into the use of a rg i l la ceous rocks and evaporites . The ob ject ive of th is research is to answer the question "Can the geologist guarantee i s o la t io n ? " posed by de M ars i ly (4) and w ith in the framework of the massive in ternat ional and national act i v i t ies in th is topic.
The present paper reviews the geological c r i t e r i a referred to above in the l igh t of subsequent in ternat ional pub lications and describes the philosophy behind the geological component of the re levant UK research programmes and th e i r present s t a tu s .
CRITERIA, DEFINITION AND REVIEW
The eight c r i t e r i a proposed for the se lec t ion of areas and referred to above are not repeated here but are summarised in Table I , together with the factors requiring ana lys is in each. The basis leading to th e i r d e f in i t io n included the fo l lowing assumptions : -
1. That only s o l id i f i e d and packaged reprocessing wastes are considered.
2. That the o b jec t ive of th e i r disposal into a host-rock is to ensureth e i r containment such that they do not return to the biosphere in amounts const i tu t ing a b io log ica l hazard and, in p a r t ieu 1ar>that they w i l l not exceed the app licab le dose l im its ( 5) .
3. That there is no intention to re t r ie v e the wastes and that su rve i l la n ce would not be maintained a f te r repos itory c losure.
!*. That containment should remain effective for a period of the orderof a hundred thousand years or so (5 ) .
5. That containment should be seen as a mu 11 i-barr ie r concept, including independent engineered and geologica l components, against the background that mobile groundwaters are the most probable d ispersal medium i f containment were to be breached.
6 . That the c r i t e r i a should be in te rn a l ly consistent and th a t th e i r mutual in te ract ions w i l l not lead to loss of containment.
7. That the consequences of the occurrence of various geological and c l im a t ic events should be id en t if ied and th e i r impact assessed.
8 . That no one s i t e w i l l meet a l l of the c r i t e r i a but that a balance appropriate to the p a r t i cu la r host-rock and geological environment would be necessary.
The in te rac t ion between these geological c r i t e r i a and numerous other te chn ica l , s o c ia l , p o l i t i c a l , e th ica l and environmental factors was outs ide the remit of the Working Party which prepared the o r ig ina l Report (2) which was published as a contr ibution to the in te rnat iona l debate on disposal to geological formations. The c r i t e r i a are now reviewed b r i e f l y in the context of a few of the many subsequent contr ibutions to the same debate, w ith p a r t i cu la r but not exc lus ive reference to those publihsed by the IAEA (5 ) ; Frye ( 6) ; G i l e t t i and S iever (7 ) ; Brederhoeft ( 8) ; Deutsch (9) and KBS (10).
IAEA-SM-243/29 457
TABLE I. AREA SELECTION CRITERIA
CRITERION FACTORS FOR ANALYSIS
1 . Gross 1i thology and spa tia l d is t r ib u t io n of rocks.
3.
Phys ica l and chemical c h a ra c t e r is t i c s in re la t io n to wastes and conta i n e rs .
Hydrological and hydro- geological criteria.
h. Seismic conditions.
5. Re la t ion to other man- made structures and poss ib le mineral deposits .
6 . Areal c h a ra c t e r is t i c s in re la t io n to possib lec l imat i с change.
7- Leakage routes to the su r f a c e .
8. Engineering practices.
Geological h is to ry of area; topography and geomorphology; rock type and homogene ity ; macro-and micro-1 ithology; petro logy; thickness and la te ra l extent; s t ruc tu ra l in t e g r i ty and s t a b i l i t y ; s ta te of s t re ss ; l i tho logy and thickness o f overburden and flanking rocks; potent ia l erosional prob1ems.Thermal properties of host rock re levant to heat d is s ip a t io n and capac i ty to withstand thermal s t re ss ; hydrau lic co ndu ct iv i ty ; potent ia l fo r physicochemical in te rac t ion between wastes.Occurrence of surface water above reposi to ry ; geomorphological s t a b i l i t y ; hydrogeological condit ions including groundwater c i r c u la t io n , chemistry and the potent ia l influence of thermal e f f e c ts .H is to r ic a l seismology; return periods for events of given seismic in te n s it y ; d i re c t and ind irec t e f fe c ts of se ism ic i ty upon repos itory .
H is to r ic a l sub-surface a c t i v i t i e s ; potent ia l for mineral reserves and resources; d i re c t and ind irec t e f fe c ts of other sub-surface a c t i v i t i e s .Repository depth in re la t io n to possib le acce lera ted rates of erosion; e f f e c ts of g l a c i a l , i n te rg1a c i a l , p luv ia l and ar id ep i sodes.Po ten t ia l flow paths from the repos itory to the surface ; sorption and/or f ix a t io n potent ia l of rocks on that flow path; possib le d i lu t io n or concentrat ion during groundwater movement.Excavation technology; excavation s t a b i l i t y ; rock physico-mechan i cal p ropert ies ; in s i tu s t ress cond it ions, w ith and without repos itory in s ta l le d , w ith and without thermal s t re ss ; repos itory sea ling technology.
1. Gross lithology and spatial distribution of the rooks
The three p r inc ipa l host rocks prev ious ly id en t if ied remain under in te rnat iona l study v iz , rock s a l t ( p r in c ip a l l y domed, but a lso bedded s a l t s ) , c r y s t a l l i n e rocks ( inc luding in t ru s iv e s , extrus ives and metamorphics) and a rg i l la ceous m ate r ia ls . The pos it ion has changed somewhat in that rock s a l t , o r i g i n a l l y the front runner, is perhaps no longer seen in qu ite that l igh t (9 , p. 2 6 ) .
458 MATHER et a l
However, there seems to be no reason to modify the o r ig ina l c r i t e r io n that " . . . . n o formation meeting the access and containment requirements need be excluded", and th is 'v ie w has received subsequent support (9 , p. 26) .
L i t t l e , i f any, fresh information has been issued on desirab le minimum dimensions for blocks of host-rock able to contain repos ito r ie s . D if fe rent factors a r ise for storage of spent fue ls but th is paper is not concerned with that problem.The need for a minimum thickness of 500m for a rg il la ceous m ater ia ls , advocated in 1976, has been questioned (Cohen, L. .p r i v a t e communication). The j u s t i f i c a t i o n remains q u a l i t a t i v e and re la tes to the low level of p r e d ic t a b i l i t y that a block of a rg i l la ceous material does, or does not, contain laminae of s i l t or sand which could act as a permeable route for mobile groundwater, of whatever o r ig in and i r re sp ec t ive of whether that groundwater is moving under regional hydrau lic gradients or convective thermal e f f e c ts .
The minimum depth of buria l of 300m advocated by Schneider and P l a t t remains la rge ly unchallenged although the IAEA (5) consider that s l ig h t l y shallower bur ia l might be permissib le in some condit ions. In that context, however,Roberts (1) has argued that the dispersal to the environment of uranium and its daughters caused by acce lera ted erosion of a gran ite to a depth of 300m and over an area of repos itory size is more l i k e l y to cause an add itiona l hazardthan s im ila r e f fe c ts on deeply buried wastes in a repos itory of the same sizea f te r 10 000 years.
2. Physical and chemical characteristics in relation to wastes and containers.
The l i t e r a tu r e on rock properties in re la t io n to container and repository design, nuclide transport and sa fe ty ana lys is has grown to a major extent since pub l ic a t ion of the c r i t e r i a in 1976- Yet, due to the operation of la rge ly non-techni c a 1fa c to rs , d isappo in ting ly few of the many new data are derived from potentia lhost rocks, other than rock-sa lt , at s i te s under a c t ive invest iga t ion for repos itory research. For example, among innumerable laboratory values for d ispers ion c o e f f ic ie n t s those measured by standard methods and capable of comparison with one another are noticeable by th e i r s c a r c i t y . S i t e sp e c i f i c data measured in s i tu are few and w i l l be d i f f i c u l t to acquire at repos itory s i te swithout damaging the s i t e in te g r i ty .
Since the c r i t e r i a were compiled the d e s i r a b i l i t y of r e s t r i c t in g the temperature of the repository to less than 100 C, c e r t a in ly less than 200°C, has been advocated by Chapman (11) andBrederhoeft ( 8) . In the same context i t isnoteworthy that several of the many in ternat iona l reviews of the Swedish KBSproposals recognise the advantage of e l im inat ing some problems by the adoption of a maximum repository temperature of 65 С (12).
S. Hydrogeological and hydrological criteria.
In th is f i e ld of research much recent e f fo r t has been d irected towards the id e n t i f ic a t io n of methods of measurement in low permeability media. P a ra l le l work being undertaken in the development of 'Hot dry rock' energy systems at Los Alamos and elsewhere is of d i re c t in te res t although the two workshops at Austen,Texas (1977) and Pa r is (1979) were d irected at waste disposal problems.In th is context the physical meaning of "p o ro s i ty " in c r y s t a l l i n e rocks such as gran ite must be examined before i t can be accepted that such m ater ia ls carry of the order of Ц of free groundwater.
The 'Hot dry rock' studies bear on two other fa c to rs of in te re s t . F i r s t , the research on the v e r t i c a l i t y of induced f is su r in g under n a tu ra l ly occurring s t re s s - f ie ld s bears on the id e n t i f ic a t io n o f potent ia l leakage paths for mobile groundwaters. This w i l l be s ig n i f ic a n t where the length of flow paths is a sen s i t iv e parameter in s i t e - s p e c i f i c sa fe ty assessments, rather than in the
IAEA-SM-243/29 459
generalised assessments of the type published by H i l l and Grimwood (13) inwhich the flow path to the surface is assumed to be 10km . Second, many dataare becoming a v a i la b le on f lu id geochemistry at re levant temperatures and pressures
As a general po int , G i l e t t i and S ieve r (7) r ig h t l y ind icate that most large g r a n i t i c in t rus ions , of the order of 2 to 15 km th ick , w i l l not be underlain by aq u i fe rs , in advantageous con trad is t in c t ion to the potent ia l sedimentary host rocks I t can be argued s im i la r l y that the plutons w i l l not have s ig n i f ic a n t aquifers l a t e r a l l y adjacent to repos ito r ies placed away from th e i r margins.
4. Seismic conditions.
The need to consider se ism ic i ty as a factor in the design of a repos itory in a region such as the UK character ised by a low seismic hazard is not se lf-ev iden t . Credence is lent to that view by the in ternat iona l review by P r a t t (14) bf earthquake damage to underground f a c i l i t i e s . Indeed the data for measured displacements as a function of depth tend to suggest that se ism ic i ty need not be a matter of s ig n i f ic a n t concern in repos itory design, although i t s tec ton ic s ig n if icance in re la t io n to regional s t a b i l i t y must continue to be recognised.
5. Relation to other man-made structures arid possible mineral deposits.
The essen t ia l need for id e n t i f ic a t io n and accurate location o f h is to r i c a l sub-surface structures has been emphasised in re la t ion to s a l t by G i l e t t i and S iever (7) and the need must be recognised in re la t io n to the other potentia l host rocks, although less c r i t i c a l l y . Frye ( 6) suggests that records of resource potent ia l should be augmented by " . . . . a dependable appraisa l of the- kinds of resources that may be present ( inc lud ing those of low grade or unorthodox p o te n t ia l ) " . In terpreted s t r i c t l y that could represent a major component of a s i t e in vest iga t ion programme, but placed in a broader context i t should probably be recognised as reasonably des irab le to a level which w i l l have to be decided on a s i te-speci f i с basis .
6. Aereal characteristics in relation to possible climatic change.
I t should be taken as axiomatic that the global c limate w i l l not remain stab le over the relevant time period. The d ire c t ion of departure from the present s ta te cannot be predicted and design c r i t e r i a should therefore incorporate features relevant to e i th e r greater a r id i t y or greater humidity, as well as to e i th e r renewed g la c ia t io n or further am eliora tion of the present c l im ate . So fa r as s i t in g c r i t e r i a are concerned G i l e t t i and S iever (7) r ig h t l y point to the advantage of a locat ion l i k e l y to su f fe r g la c ia t io n because of the concomitant is o la t ion " . . . f r o m intrusion (by mankind) several thousands of years hence". A somewhat s im i la r argument can be advanced in terms of submergence of the land due to an advance of the sea fo l lowing melting of polar ice. On that basis the o r ig ina l c r i t e r io n to se lec t a s i t e at an e levat ion greater than 60m above mean sea level and so escape submergence can be relaxed.
7. Leakage routes to the surface.
Theoretica l modelling studies of pathways back to man have advanced considerably in the last three years, but real data re la t ing to sp e c i f i c s i te s for which the models can be tested remain lim ited to a few components. Recognising the problems of c o l le c t in g many such data in a meaningful manner and in s t a t i s t i c a l l y s ig n i f ic a n t numbers, the proposal by H i l l and Webb (15) that the range of data requirements should be reduced by s e n s i t i v i t y ana lys is is welcome.
460 MATHER et al.
8 . Engineering p m r t v » . ' .
Re c e n t a d v a n c e s have been m a d e in r e p o s i t o r y des i g n in relation to both
a r g i l l a c e o u s and c r y s t a l l i n e rocks (161 to add to an a l r e a d y e x t e n s i v e litera t u r e
on de s i g n features for a salt repository. Some of the p r o p o s a l s a d v a n c e d in
the UK for r e p o s i t o r y d e s i g n s (17), w h i c h depart in p r i n c i p l e from the IGS
c ri t e r i a , arc interesting in concept but ap p e a r to the p r esent a u t h o r s p r o b l e m a t i c
as to lon g - t e r m c o n t a inment.
SITE SELECTION
The a p p l i c a t i o n o f the g e o l o g i c a l c r i t e r i a r e s u l t e d in the r e c o g n i t i o n o f o v e r 100 a r e a s c o n t a i n i n g p o t e n t i a l l y s u i t a b l e h o s t r o c k s ( c r y s t a l l i n e i gn e o u s and meta mor ph ic r o c k s , a s w e l l a s a r g i l l a c e o u s and e v a p o r i t e f o r m a t i o n s ) . Th es e r ange i n a r e a l e x t e n t f rom l e s s than 5 sq km to a r o u nd 6 000 sq km, t he mean b e i n g a r o u n d 260 sq km. The d i s t r i b u t i o n o f the a r e a s i n v o l v e d I s shown in Chapman, Gray and M a t h e r (18) and c o v e r s n e a r l y 39 000 sq km, a p p r o x i m a t e l y 16% o f the l a nd a r ea o f the U n i t e d Kingdom.
In o r d e r t o s e l e c t r e p r e s e n t a t i v e a r e a s f o r f e a s i b i l i t y s t u d i e s t h e y were d i v i d e d a s a d e s k s t u d y u s i n g the p r i m a r y f a c t o r s o f g e o l o g i c a l e n v i r o n me n t and e v o l u t i o n a r y h i s t o r y . The p r i m a r y g r o u p s t hu s o b t a i n e d were then f u r t h e r d i v i d e d on the b a s i o f c o m p o s i t i o n a l d i f f e r e n c e s , de g r ee o f c o n s o l i d a t i o n ( f o r a r g i l l a c e o u s r o c k s ) , s t r u c t u r a l c o m p l e x i t y and o t h e r f a c t o r s t o p r oduc e a t e n t a t i v e o r d e r o f p r e f e r e n c e b ased e n t i r e l y on g e o l o g i c a l f a c t o r s . C l e a r l y the d a ta a v a i l a b l e f o r s u ch d e s k s t u d i e s i s v a r i a b l e and i s dependent upon t he p r e s e n c e o r a b se n ce o f a p p r o p r i a t e r e c en t g e o l o g i c a l r e s e a r c h r e s u l t s . From a g e o l o g i c a l v i e w p o i n t f i e l d s t u d i e s wou ld be e s s e n t i a l b e f o r e t he o r i g i n a l s e l e c t i o n s can be c o n f i r m e d o r r e j e c t e d . However , many o t h e r f a c t o r s , i n c l u d i n g l a nd o w n e r s h i p and a c c e s s , e n t e r i n t o c o n s i d e r a t i o n at t h i s s t a g e and must be t ake n i n t o a c c o u n t in t he f i n a l s e l e c t i o n . The d e s k s t u d i e s r e s u l t e d in the c h o i c e o f a number o f a r e a s , e a c h c h a r a c t e r i s e d by a d i f f e r e n t g e o l o g i c a l e n v i r o nm e nt f o r d e t a i l e d r e s e a r c h i n t o t he f e a s i b i l i t y o f g e o l o g i c a l d i s p o s a l .
In the c r y s t a l l i n e meta mor ph i c and i g n e o u s r o c k s i t i s i n t e nd ed t o r e l a t e the p r i m a r y f a c t o r , g r o u n d w a t e r f l o w , to r oc k mass and r oc k m a t e r i a l p a r a m e t e r s , s uch a s m i n e r a l o g y and p e t r o l o g y , j o i n t f r e q u e n c y , a t t i t u d e and o p e n n e s s and r e g i o n a l s t r e s s e s and t o b u i l d up a d a ta base s o t h a t the most a p p r o p r i a t e c r y s t a l l i n e e n v i r o n m e n t s f o r c o n s i d e r a t i o n a s r e p o s i t o r y s i t e s can be d e f i n e d .
E xampl es o f the t y p e s o f g e o l o g i c a l e n v i r o n m e n t c ho se n f o r f e a s i b i l i t y s t u d i e s a r e a s f o l l o w s ( not in o r d e r o f p r i o r i t y ) :
1. C a l e d o n i a n g r a n i t e i n t r u d e d i n t o (1A) Lower P a l a e o z o i c s e d i m e n t so r ( I B ) v o l c a n i c r o c k s . I n t r u s i o n 1A i s a zoned body w i t h an o u t e r u n i t o f b i o t i t e t o n a l i t é m er g i n g p r o g r e s s i v e l y i n t o a c e n t r a l u n i t o f b i o t i t e g r a n i t e . I t i s i n t r u d e d i n t o O r d o v i c i a n s e d i m e n t s o f v a r i o u s t y p e s , p r e d o m i n a t e l y g r e y w a c k e s , but w i t h s i g n i f i c a n t d e v e l op me n t s o f b l a c k s h a l e s and c h e r t s and has a s h a r p c o n t a c t w i t h an e x t e n s i v e t hermal a u r e o l e . I n t r u s i o n IB v a r i e s in c o m p o s i t i o n f rom a py r ox en e g r a n i t e t o a p y r o x e n e - f r e e g r a n o p h y r e , i s i n t r u d e d i n t o a n d e s i t e l a v a s and has a f a u l t e d b r e c c i a t e d c o n t a c t .
2. C a l e d o n i a n g r a n i t e f o r c e f u l l y i n t r u d e d i n t o P re - C a m b r i a n metamor ph i c r o c k s . T h i s i n t r u s i o n i s a c o a r s e h o r n e b l e n d e - b i o t i t e t o n a l i t é v a r y i n g t ow ar ds the o u t e r p a r t t o g r a n o d i o r i te. A number o f SW-VlE
IAEA-SM-243/29 461
t r e n d i n g f r a c t u r e s c r o s s the g r a n i t e w h i ch i s i n t r u d e d i n t o metamorphosed s e d i m e n t s b e l o n g i n g m a i n l y to the P r e - C a m b r i a n Mo ine S u p e r g r o u p .No thermal a u r e o l e i s d ev e l o p e d but the i n t r u s i o n m ar g in i s g e n e r a l l y i n t i m a t e l y we lded to the c o u n t r y r oc k by the de ve lopment o f c o n t a c t mi gmat i t e s .
3. C a l e d o n i a n g r a n i t e p e r m i s s i v e l y i n t r u d e d i n t o P r e - C a m b r i a n metamor ph i c r o c k s . I n t r u s i o n 3 i s a r i n g complex composed o f 4 s u c c e s s i v e g r a n i t e i n t r u s i o n s emplaced a s a r e s u l t o f p r o g r e s s i v e c a u l d r o n s u b s i d e n c e s . The f a v o u r e d i nnermost u n i t i s a q u a r t z a d e m e l l i t e w h i ch has a s i m p l e n e a r - v e r t i c a l c o n t a c t and i s d y k e - f r e e .
b. G r a n i t i c compl ex in the c o r e o f a r e g i o n a l m i g m a t i t e z one. I n t r u s i o nh o c c u p i e s the c o r e o f an i n j e c t i o n complex w i t h i n m e ta s ed i me n ts o f the M o i ne S u p e r g r o u p . The i n j e c t e d m a t e r i a ) i s d o m i n a n t l y b i o t i t e g r a n i t e o f v a r i a b l e g r a i n s i z e w h i ch o c c u r s as v e i n s o r c o n c o r d a n t s h e e t s w i t h i n the p s a m m i t i c h o s t .
5. C a l e d o n i a n b a s i c i n t r u s i v e . I n t r u s i o n 5 i s an o l i v i n e - g a b b r o w i t h a s s o c i a t e d c u m u l a t e - d e r i v e d p e r i d o t i t e s and c o n t a m i n a n t - d e r i v e d n o r i t e s . I t i s s t e e p - s i d e d w i t h s t e e p e r d i p s on the west than on the e a s t , i m p l y i n g an i n c r e a s e in w i d t h a t d ep t h . Much o f the i n t r u s i o n i s e x t e n s i v e l y a l t e r e d w i t h p y r o x e ne r e p l a c e d by a mp h ib o l e .
6. N o n- m i g m a t i sed P re - C a m b r i a n Mo ine M e t a s e d i m e n t s . The r o c k s o f the Mo ine S u p e r g r o u p c o n s i s t o f t h i c k metamorphosed a r e n a c e o u s and a r g i l l a c e o u s s e d i m e n t s w i t h some i n t e r m e d i a t e t y p e s . A re a 6 has been c ho se n w i t h i n a r e g i o n o f m a s s i v e t o f l a g g y q u a r t z o - f e I d s p a t h i c p sammi tes o f i n t e r m e d i a t e meta mor ph ic g r a d e and min imal s t r u c t u r a l d i s t u r b a n c e .
7. P r e - C a m b r i a n L e w i s i a n basement r o c k s . The L e w i s i a n i n c l u d e s r e l i c s o f s e v e r a l o r o g e n i c c y c l e s c o v e r i n g a t ime span o f a t l e a s t 2800X 1 0 6a t o the l a t e s t r e w o r k i n g at 1 6 0 0 X 1 0 6 a. The major s u b - d i v i s i o n i s i n t otwo t i m e - b a s e d c om pl e xe s , the o l d e r S c o u r i a n and the y o un g e r L a x f o r d i a n . A re a 7 has been c ho s en w i t h i n an a r e a o f L a x f o r d i a n r o c k s w h i ch a r e d o m i n a n t l y g r e y q u a r t z o - f e l d s p a t h i c and b i o t i t e h o r n b l e n d e g n e i s s e s w i t h i n t r u s i v e s h e e t s and v e i n s o f g r a n i t e and p e g m a t i t e . The g n e i s s e s have und er g on e t h o r o u g h r e - w o r k i n g r e s u l t i n g in r o c k u n i t s t h a t a r e r e l a t i v e l y homogeneous compared w i t h much o f the r e s t o f t he L e w i s i a n b a s e m e n t .
A c c e s s i s now needed t o r e s e a r c h s i t e s in w h i ch t o mount the f e a s i b i l i t y s t u d i e s . B e f o r e work can p r oce ed a p p l i c a t i o n s have t o be s u b m i t t e d under the p l a n n i n g r e g u l a t i o n s f o r p e r m i s s i o n t o d r i l l i n v e s t i g a t o r y b o r e h o l e s . So f a r s uch a p p l i c a t i o n s have been s u b m i t t e d f o r s i t e s in a r e a s 1A, I B and к but o n l y t ha t w i t h i n a r e a b has been g r a n t e d . T h i s a r e a c o m p r i s e s the S t r a t h H a l l a d a l e I n j e c t ion Complex and l i e s in n o r t h - e a s t S c o t l a n d . The r e s e a r c h s i t e o c c u p i e s some •100 sq km w i t h i n C a i t h n e s s D i s t r i c t and i s peat and d r i f t c ov e r e d w i t h o n l y l i m i t e d e x p o s u r e o f be dr oc k . A p r e l i m i n a r y d r i l l i n g programme has been d e s i g n e d t o c o n f i r m the l i m i t e d g e o l o g i c a l i n f o r m a t i o n and 3 b o r e h o l e s have been d r i l l e d to 300m, t o g e t h e r w i t h 2b b o r e h o l e s to some ^Om. The da ta o b t a i n e d f rom the b o r e h o l e s complemented by f i e l d mapp ing, s u r f a c e and b o r e h o l e g e o p h y s i c s and d e t a i l e d h y d r o g e o l o g i c a l measurement s a r e b e i n g used in a p r e l i m i n a r y r ev i ew of. the c o n t a i n me n t c h a r a c t e r i s t i c s o f t h i s p a r t i c u l a r g e o l o g i c a l e n v i r o n m e n t .A d e c i s i o n on w he th er o r not t o d r i l l f u r t h e r b o r e h o l e s to s t u d y the p r o p e r t i e s o f the formation at p o t e n t i a l r e p o s i t o r y d ep t h s w i l l be made o nce t h i s r e v i e w has been compl eted .
462 MATHER et al.
The a p p r oa c h used in the s t u d i e s o f the f e a s i b i l i t y o f g e o l o g i c a l d i s p o s a l to a r g i l l a c e o u s r o c k s and e v a p o r i t e s i s b r o a d l y c ompa ra b l e to t ha t used
f o r the c r y s t a l l i n e r o c k s . The f a c t o r p a r t i c u l a r l y r e l e v a n t to the s e l e c t i o n o f t h e s e a r e a s i s the p r e d i c t a b i l i t y o f t h e i r mass c h a r a c t e r i s t i c s . I m p e r s i s t e n t de ve lo pm en t s o f m i no r l i t h o l o g i c a l v a r i a t i o n s may not pos e p r ob l em s but the p r e s e n c e o f , f o r i n s t a n c e , a t h i n c a l c a r e o u s s a n d s t o n e w i t h i n an o t h e r w i s e homogeneous g r e y w a c k e / s i 1t s t o n e s equ ence , c o u l d be o f m aj o r s i g n i f i c a n c e f r om the v i e w p o i n t o f w as t e c on t a i n me n t . A u s e f u l p r o p e r t y o f p l a s t i c c l a y s i s t h e i r c a p a c i t y f o r s e l f - s e a l i n g . Wi th p r o g r e s s i v e c om pa ct i on and b u r i a l t h i s p r o p e r t y w i l l be l o s t but w i l l be c o u n t e r e d by an i n c r e a s e in thermal s t a b i l i t y a s a r e s u l t o f the e x p u l s i o n o f n o n - s t r u c t u r a l l y h e l d w a t e r . As t h es e c h a n ge s a r e m u t u a l l y e x c l u s i v e an i n t e r m e d i a t e compromi se ha s g e n e r a l l y been s o u g ht i n v o l v i n g f o r I n s t a n c t s e l e c t i o n o f a r e a s c o n t a i n i n g u n c l e a v e d m uds ton es .
A l t h o u g h s a l t domes o c c u r b en ea th the UK C o n t i n e n t a l S h e l f t hey a ppe ar to be a b s e n t b e ne a th the l andmas s and t he e v a p o r i t e f o r m a t i o n s a v a i l a b l e a r e a l l bedded d e p o s i t s . Th es e g e n e r a l l y o c c u r w i t h i n t h i c k e r a r g i l l a c e o u s f o r m a t i o n s and a r e h y b r i d s e qu e nc es c o n t a i n i n g i n t e r b e d d e d e v a p o r i t e s and muds tones and / or s i l t s t o n e s .
T y p i c a l t y p e s o f a r g i l l a c e o u s and e v a p o r i t e f o r m a t i o n s e l e c t e d f o r more d e t a i l e d s t u d a r e the f o l l o w i n g :
1 A s e r i e s o f g e n t l y f o l d e d O r d o v i c i a n and S i l u r i a n r o c k s do mi na t ed by a r g i l l a c e o u s t y p e s t o t a l l i n g some 1(000 m in t h i c k n e s s . The r o c k s a r e w e l l e x p os e d a t the s u r f a c e and d i p c o n s i s t e n t l y t o the s o u t h - e a s t at an a n g l e o f ab ou t **5°.
2 A s equence o f t u r b i d i t i c m uds ton es w i t h i n t e r b e d d e d s i l t s t o n e s and l i m e s t o n e s o f Lower C a r b o n i f e r o u s age. The r o c k s o c c u r w i t h i n af a u l t - b o u n d e d t r o u g h and a minimum t h i c k n e s s o f 1000 m o f a r g i l l a c e o u s r o c k s i s a v a i l a b l e . These r o c k s c o n s t i t u t e a good example o f the e f f e c t i v e compromi se d e s i r a b l e in a r g i l l a c e o u s r o c k s - a t h i c k f o r m a t i o n t h a t h a s been dewat ered and d e v e l o p e d a mature c l a y m i n e r a l o g y by deep b u r i a l , w i t h o u t the a t t e n d a n t de ve lopment o f s t r u c t u r a l c o m p l e x i t y .
3 A m u d s t o ne - do m i na t ed s equence o f T r i a s s i c r o c k s a t t a i n i n g t h i c k n e s s e s j u s t i n e x c e s s o f 500 m. Th es e o c c u r w i t h i n the e v a p o r i t e f r e e p a r t o f a major d e p o s i t i o n b a s i n w i t h a c o v e r o f up t o 500 m o f m a i n l y a r g i l l a c e o u s r o c k s o f J u r a s s i c age.
A s e r i e s o f S i l u r i a n muds tones and g r e y w a ck e s w i t h an o v e r a l l t h i c k n e s s o f a r ou nd 5000 m. Th es e r o c k s a r e f o l d e d on bo th a l a r g e and s ma l l s c a l e and a r e c l e a v e d t o a l e s s e r o r g r e a t e r e x t e n t d e p e n d i n g on l i t h o l o g i c a l v a r i a t i o n s .
5 H y b r i d e v a p o r i t e / m u d s t o n e s e qu e nc es w i t h i n P e r m o - T r i a s s i c s e d i m e n t sd ev e l o p e d in m aj o r d e p o s i t i o n a l b a s i n s . Thes e h y b r i d s e d i m e n t s show 3 o r d e r s o f c y c l i c i t y v a r y i n g f rom mm s c a l e t o 1 0 0 ' s o f m e tr e s .The r o c k s o f i n t e r e s t w i t h i n the b a s i n s a r e in e x c e s s o f 600m in t h i c k n e s s and a r e g e n e r a l l y o v e r l a i n by o t h e r a r g i l l a c e o u s f o r m a t i o n s . In a l l t h e s e e xamples e x a m i n a t i o n o f the c y c l i c i t y o f the h y b r i d s equences f rom t he p o i n t o f v ie w o f w a s te i s o l a t i o n i s a p a r t i c u l a r l y i mpo rta nt a s p e c t .
I t i s hoped t o mount s t u d i e s t o examine the f e a s i b i l i t y o f g e o l o g i c a l d i s p o s a l t o a r g i l l a c e o u s r o c k s a t each t ype o f s i t e . Work has not a dvanced a s f a r as w i t h the c r y s t a l l i n e r oc k s t u d i e s and a t t he p r e s e n t t ime no a p p l i c a t i o n s f o r p e r m i s s i o n to d r i l l b o r e h o l e s have been s u b m i t t e d .
IAEA-SM-243/29 463
CONCLUSION
The p r i n c i p a l c o n c l u s i o n w h i ch ap pe ar s to emerge f rom the m u l t i - d i s c i p l i n a r y e f f o r t s o f t he l a s t t h r e e y e a r s , i s the need f o r v i g o r o u s p u r s u i t o f c o m p r e h e n s i v e but s i t e - s p e c i f i c f i e l d d a ta in a r ange o f v a r i a b l e s f rom many d i f f e r e n t g e o l o g i c a l e n v i r o n m e n t s . The f e a s i b i l i t y s t u d i e s s u g g e s t e d above wou ld p e r m i t the c o l l e c t i o n o f s uch d a t a and a l l o w f u r t h e r d e c i s i o n s t o be made on s i t e s e l e c t i o n w h i l s t t e s t i n g the o r i g i n a l c r i t e r i a more c o m p r e h e n s i v e l y .G iv en s u c h a da ta ba se the p r o s p e c t wou ld be l e s s e n e d o f t ime b e i n g l o s t u n n e c e s s a r i l y i f the f r o n t - r u n n e r in any p a r t i c u l a r programme has to be abandoned f o r u n p r e d i c t a b l e e n v i r o n m e n t a l r e a s o n s .
ACKNOWLEDGEMENT
Part of the site s e l e c t i o n p r o c e d u r e s and of the w o r k on c r y s t a l l i n e rocks has
been s u p p o r t e d und e r a c o n t r a c t b e t ween the C o m m i s s i o n of the E u r o p e a n C o m m u n i t i e s
and the U K A E A and the rem a i n d e r by the UKAEA; this s u p port is g r a t e f u l l y
a c k n o w l e d g e d . The p a p e r is p u b l i s h e d w i t h the p e r m i s s i o n of the Director,
Institute of G e o l ogical S c i e n c e s (Natural E n v i r o n m e n t R e s e a r c h Council).
REFERENCES
(1) ROBERTS, L . J . , " R a d i o a c t i v e w a s t e : p o l i c y and p e r s p e c t i v e s " . B r i t i s h N u c l e a r E n e r g y S o c i e t y 1978.
(2) GRAY, D.A . ( C h a i rman), e t a l . , " D i s p o s a l o f h i g h l у - a c t i v e , s o l i d r a d i o a c t i v e w a s t e s i n t o g e o l o g i c a l f o r m a t i o n s - r e l e v a n t g e o l o g i c a l c r i t e r i a f o r the U n i t e d K i n gd om " , 1976, Rep. I n s t . G eo l . S c i . , No. 76/12. HMSO, London.
(3) COMMAND, 681S , N u c l e a r Power and t he E n v i r o n m e n t , R oya l Co mm is s i o n on E n v i r o n m e n t a l Pol l u t i o n , ' S i x t h R e p o r t , 1976, HMSO, London.
(4) DE M A RS I L Y , e t a l . , N u c l e a r w a s t e i s o l a t i o n : Can the g e o l o g i s t g u a r a n t e ei s o l a t i o n ? S c i e n c e 197 (1977) 519-
( ? ) I A E A , " S i t e s e l e c t i o n f a c t o r f o r r e p o s i t o r i e s o f s o l i d h i g h - l e v e l and a l p h a b e a r i n g w a s t e s in g e o l o g i c a l f o r m a t i o n s " , 1977, T e c h n i c a l R epo rt S e r i e s No. 177.
(61 FRYE, J . C . (Cha i rnan) , et a l . , " G e o l o g i c a l c r i t e r i a f o r r e p o s i t o r i e s f o rh i g h - l e v e l r a d i o a c t i v e w a s t e s " , 1978, Panel on G e o l o g i c a l S i t e C r i t e r i a . Committee on R a d i o a c t i v e Waste Management. N a t i o n a l Academy o f S c i e n c e s .
(7) G I LETT I , B. and S I E V E R , R. ( Co-Cha i rmen)> e t a l . , " S t a t e o f g e o l o g i c a lk nowledge r e g a r d i n g p o t e n t i a l t r a n s p o r t o f h i g h - l e v e l r a d i o a c t i v e w as te from deep c o n t i n e n t a l r e p o s i t o r i e s " , 1978, R epo rt o f an Ad Hoc Panel o f E a r t h S c i e n t i s t s , E n v i r o n m e n t a l P r o t e c t i o n A g e n cy , W a s h i n g t o n .
(5) BREDERHOEFT, J . D . , e t a l . , " G e o l o g i c d i s p o s a l o f h i g h - l e v e l r a d i o a c t i v e w a s t e s - e a r t h s c i e n c e p e r s p e c t i v e s " , 1978, U S G e o l o g i c a l S u r v e y C i r c u l a r
779.
464 MATHER et al.
(У) DE'ITSCH, J.M. (Chairman) , Report to the P r e s i d e n t by the Inter a g e n c y
R ev iew Group on N u c l e a r Waste Management, 1978, T I D - 28817 ( D r a f t ) and TOD - 28818 ( D r a f t ) , Depar tment o f E n e r g y , W a s h i n g t o n .
(10) K B S , " H a n d l i n g o f s p e n t n u c l e a r f u e l and f i n a l s t o r a g e o f v i t r i f i e dh i g h - l e v e l r e p r o c e s s i n g w a s t e " , 1978, 5 v o l ume s, K B S P r o j e c t , S t o c k h o l
(111 C HAP M A N , N.A., M i n e r a l o g i c a l and g e o ch e mi c a l c o n s t r a i n t s on maximum
p e r m i s s i b l e r e p o s i t o r y t e m p e r a t u r e s , these Proceedings, SM-243/28.
(.12) I NDUSTRI DEPARTEMENTET. , " R e p o r t on r e v i e w t h r o u g h f o r e i g n e x p e r t i s e o f the r e p o r t h a n d l i n g o f s p e n t n u c l e a r f u e l and f i n a l s t o r a g e o f v i t r i f i e d h i g h l e v e l r e p r o c e s s i n g w a s t e " , DSI 1 9 7 8 : 2 8 , S to ck ho lm.
(13) H I L L , M.D. and GRIMVOOD, P . D . , " P r e l i m i n a r y a s s e s s m e n t o f the r a d i o l o g i c a l p r o t e c t i o n a s p e c t s o f d i s p o s a l o f h i g h l e v e l w as te in g e o l o g i c f o r m a t i o n s " . 1978, N R P B - R - 6 9 •
(141 PRATT, H.R. et a l . , " E a r t h q u a k e damage to u n de r g r ou n d f a c i l i t i e s . "D P - 1 5 1 3 , I 9 7 d , Sava nn ah R i v e r L a b o r a t o r y , A i k e n , S o ut h C a r o l i n a .
(15) HIL.L, M.D. and WEBB, G . A . M . , A p p l i c a t i o n o f the r e s u l t s o f r a d i o l o g i c a l a s s e s s m e n t s o f h i g h - l e v e l was te d i s p o s a l , these Proceedings, SM-243/25.
(16) BERGMAN, M. ( E d . ) . , " S t o r a g e in e x c a v a t e d r ock c a v e r n s " . R o c k s t o r e 1977, ( Pr oc . 1s t I n t . S ympo s i um) , V o l . 3 - , S to ck ho lm .
(17) G R I F F I N , J . R .j e t a l . , Geo log i ca 1 d i s p o s a I o f h i g h - l ev e I rad i o a c t i ve w as te c o n c e p t u a l r e p o s i t o r y d e s i g n in har d r o c k , these Proceedings, S M - 243/43.
(18) CHAPMAN, N.A., e t a l . , " N u c l e a r w as t e d i s p o s a l ; the g e o l o g i c a l a s p e c t s " .New S c i e n t i s t , 78 ( 1 9 7 8 ) , 225.
DISCUSSION
O. STEPHANSSON: In the paper you list seven examples of the types of
geological environment chosen for feasibility studies. If you had access to all
these environments, which one would be your first choice?
J.D. MATHER: At present the aim of the United Kingdom is to provide
a data base which will enable such a choice to be made. In our present state of
knowledge it is not practicable to place these geological environments in any
order of priority.
O. STEPHANSSON: The data yielded by the Swedish programme indicate
that the young, late-kinematic granites are the most fractured and the least
healed granitic rock type and therefore the most permeable. In the list of geo
logical environments you suggest I would recommend the Precambrian Lewisian
basement rock which has been subjected to several orogenic cycles.
IAEA-SM-243/29 465
J.D. MATHER: We recognize that parts of the Lewisian, particularly the
Laxfordian, should be more annealed than the Caledonian granites and therefore
less permeable. However, the Lewisian basement rocks have suffered considerable
structural deformation and have been intruded by Tertiary dykes since the last
metamorphic event and the situation found in the United Kingdom may not be
strictly comparable to that within the Scandinavian Shield.
L.J. ANDERSEN: Am I right in interpreting your comments on the
criterion relating to climatic changes to mean that you do not require an eleva
tion higher than 60 m above sea level for the entrance to a repository?
J.D. MATHER: The original criterion in relation to a polar ice sheet was
defined by recognizing the potential disadvantages of marine submergence of
the entrance to a repository if it were to be placed below an elevation of 60 m
above mean sea level. Although the associated problem of effective sealing of
the repository entrance still remains, this is considered to be offset by the
advantages due to the isolation from human intrusion which would result from
submergence.
IAEA-SM-243/93
GEOLOGICAL DISPOSAL OF
HIGH-LEVEL RADIOACTIVE WASTE
Conceptual repository design in hard rock
H. BEALE
Atomic Energy Research Establishment,
Harwell, Didcot, Oxon.
J.R. GRIFFIN, J.W. DAVIES
United Kingdom Atomic Energy Authority,
Risley, Warrington, Cheshire
W.R. BURTON
British Nuclear Fuels Ltd,
Risley, Warrington, Cheshire,
United Kingdom
Abstract
GEOLOGICAL DISPOSAL OF HIGH-LEVEL RADIOACTIVE WASTE: CONCEPTUAL REPOSITORY DESIGN IN HARD ROCK.
The paper gives an interim report on UK studies on possible designs for a repository for vitrified high-level radioactive waste in crystalline rock. The properties of the waste are described and general technical considerations of consequences of disposal in the rock. As an illustration, two basic designs are described associated with pre-cooling in an intermediate store. Firstly, a ‘wet repository’ is outlined wherein canisters are sealed up closely in boreholes in the rock in regions of low groundwater movement. Secondly, a ‘dry repository’ above sea level is described where emplacement in tunnels is followed by a loose backfill containing activity absorbers. A connection to deep permeable strata maintains water levels below emplacement positions. Variants on the two basic schemes (tunnel emplacement in a wet repository and in situ cooling) are also assessed. It is concluded that all designs discussed produce a size of repository feasible for construction in the UK. Further, (1) a working figure of 100°C per maximum rock temperature is not exceeded, (2) no insuperable engineering problems have so far been found, though rock mechanics studies are at an early stage; (3) it is not possible to discount the escape of a few long-lived ‘man-made’ isotopes. A minute increment to natural activity in the biosphere may occur from traces of uranium and its decay chains; (4) at this stage, all the designs are still possible candidates for the construction o f a UK repository.
1. INTRODUCTION
Currently the UK is storing the highly active waste liquor, which is produced during fuel processing, in cooled storage tanks.It is proposed to convert the waste into borosilicate glass contained in stainless steel canisters. There are several possible routes
467
468 BEALE et al.
for final disposal. This paper describes the studies being carried out in the UK on one option, a repository located in crystalline rock. A number of designs are being taken to the stage where they can be compared in terms of cost and safety.
2. BASIC DESIGN DATA
The repository is being designed to -accommodate 3 500 canisters forecast to arise from the UK Nuclear Programme to year 2000 AD.The radioactive characteristics and associated heat output of the waste are conveniently described by the age of the radioactive species after removal of the irradiated fuel elements from a reactor.
The canister data is as follows:
i Glass, diameter A50 mm, length 2000 mm, volume 0.32 m3.
ii Primary Sheath, length 3000 mm, thickness 10 mm.
iii Overcan1, diameter 720 mm, length 3600 mm.
iv Fission product and actinide content as oxide in glass, 15%
v Glass thermal rating, at 10 years 3.3 klf, 50 years 1.2 kW, 100 years 0.42 kW .
3. GENERAL TECHNICAL CONSIDERATIONS
During the first decade, the short-lived fission products decay away so that the canisters can be handled without forced cooling. In the period from 10 years to 100 years the radiation and heat contents of a canister originate predominantly from Sr90 and Csl37 with half-lives of approximately 30 years. After the first century the bulk of the heat output comes from Am241, half-life 430 years: this has lower energy gammas than Sr90 and Csl37, so that external radiation through the container wall continues to fall with a 30 year half-life.
Radiation and heat outputs become small after the first thousand years (conveniently described at the "short-term period"). The residual problem is then associated with the radiotoxicity of long-lived isotopes eg Am241, Am243, traces of plutonium (lost to active liquors during processing) Tc99 and Np237. In the shortterm period after emplacement, the main repository design requirement is, therefore, to establish safe thermal conditions and integrity of the engineered barriers. A maximum rock temperature of about
1 Overcanning provides extra corrosion barriers in the short term.
IAEA-SM-243/93 469
100°C has been selected for both geochemistry and civil engineering reasons. (1) The general repository size required to accept the thermal loads of the packages can be substantially reduced by cooling during the first century.
Cooling can be provided in two ways:
i in an intermediate storage position with subsequent transfer to a disposal point.
ii in the disposal position with eventual withdrawal of air flow.
Though there are general thermal stress problems which can continue after this for several centuries, the main problem becomes the adequate prevention of transfer of residual activity by ground water movement either naturally occuring'or induced by thermal perturbations. Inhibition of such a transfer can be provided in several ways:
i siting in low-permeability and highly uniform rock with low hydraulic gradients as may be found at depth or under the sea.
ii inhibiting water accumulation near packages by providing connections to permeable strata possibly under the sea.
iii delaying the release of leached activity by placing absorbing minerals in the water route back to the biosphere.
Arising from the above considerations there are a number of options in siting, depth, emplacement engineering and cooling.By way of illustration two concepts are described with the canister environments wet and dry after sealing. The effect of changes in the other parameters above is discussed later.
4. WET REPOSITORY
The basic premise of this design is that a site with minimum ground water movement is selected and canisters are sealed into the rock. The term wet is used in the sense that ground water will contact packages.
Fig.la shows the below-sea-level bore hole repository arrangement using access shafts, although an adit system is an alternative. Fig.2 is a typical loading arrangement; the inset shows deposition of the canisters in the vertical holes.
Construction underground consists of a system of transport galleries forming a grid with 20 m separation between each.
470 BEALE et al.
( a ) ( b )
DISPOSAL IN VERTICAL DISPOSAL IN HORIZONTALBORE HOLES TUNNELS
FIG .l. Deep repository (wet emplacement).
FIG.2. Vertical bore repository and transfer arrangement.
IAEA-SM-243/93 471
Emplacement holes would be drilled or raise-bored to a depth of up to 150 m and set 20 m apart.
The canisters will be emplaced such that the residual heat generated by radioactive decay is dissipated by thermal conduction, using the rock mass as a heat sink. In order to reduce the repository size (currently 450 m x 450 m x 150 m deep) it is proposed to cool the canisters in an intermediate store before emplacement for a period of up to, say, 100 years. The design utilises a three-dimensional canister emplacement configuration to reduce the amount of excavation necessary and lateral extent of the repository, which may well be an important factor in a site-specific situation.
The canisters will be grouted to improve heat transfer and offer some protection against corrosion.
The materials encasing the vitrified waste are chosen to retain the activity during the short-term period. Thereafter, the return of the radionuclides to the biosphere will be delayed by the low leach rate of the glass and the subsequent tortuous diffusion path through formations with significant absorption capabilities for most of the radionuclides involved. The long half-lives of some of the radionuclides mean that reliance must ultimately be placed on the natural geologic barrier.
5. DRY REPOSITORY
The basic premise in this concept is that ground water is directed through a series of activity absorbers eventually to displace saline water in permeable strata under the sea. The general arrangement is shown in Fig.3. Emplacement tunnels,4 m diameter, 250 m long, connect to 5 m diameter- transport galleries at enlarged T-junction6. The canisters are emplaced end to end horizontally at 4 m pitch in two rows, resting on a loose fill of minerals chosen to absorb radioactivity. A typical system of transfer from a transport gallery to emplacement tunnel is shown in Fig.4. Man access for canister emplacement within tunnels is not a requirement.
The flasl^ with its conventional end-loading arrangement can be convectively cooled during transit. It is mounted on a motorised trolley in the transport gallery, then turned through 90° horizontally to allow discharge of the canister through shielded doors into the emplacement tunnel. A remotely operated machine, the principles of which are well-proven, transfers the canister to a location on the bed of minerals adjacent to its predecessor. The design allows for retrievability of the transport machine in the event of untoward failure and a means of withdrawing the equipment to the transport gallery for
472 BEALE ET
FIG.3. Above-sea repository (dry emplacement).
FIG.4. Tunnel repository and transfer arrangement.
IAEA-SM-243/93 473
maintenance. Backfilling the tunnels and access galleries is completed with more absorber minerals. The main adits to the surface are then sealed with an impermeable backfill.
The repository is situated above sea level to allow water penetrating the tunnel walls to pass through the backfill and additional common absorber beds to deep connections to permeable strata. The displacement of saline water in the permeable strata deep under the seabed gives a long liquid holdup: activity holdup is enhanced by the absorbent properties of the strata. The high permeability of the latter induces only a small head of water above sea level in the connection passages, well below the canister emplacement level.
The 28 emplacement tunnels, at a pitch of 50 metres to reduce thermal interactions, occupy an area of roughly 0.4 km2. Maximum rock temperatures are calculated as about 75°C,giving a measure of flexibility in closer tunnel spacing or shorter emplacement time. In the short term, there is sufficient temperature drop between canister and rock to prevent liquid contact with the canisters. It is expected that the connecting systems to the permeable strata will operate satisfactorily at least in the short term. Possible flooding in the long term reduces the barrier to the surrounding local absorbers and 300 m of rock to the surface.
6. OTHER DESIGN VARIANTS
There are a number of design options at an early stage of investigation which are mentioned briefly below.
6.1 Tunnel emplacement in a deep wet repository
One variant to that described in Section 4 is to emplace in tunnels as in Fig.4 and backfill around the canisters with absorbing minerals and grout to form an impermeable matrix.The water movement in the repository after sealing then becomes similar to the wet repository described in Section 4. Fig lb shows the general arrangement. Possible advantages over the borehole system are lower construction cost and an additional "engineered" barrier.
6.2 In Situ Cooling
An intermediate long-term cooled store may be avoided if the emplacement positions in the repository receive a cooling air flow.A practical arrangement comprises tunnel emplacement as discussed in Section 5. A series of chimneys connecting the transport galleries and emplacement tunnels with the surface (see Figs 1 and 3.) provide a natural draught sufficient to draw air along the tunnels
474 BEALE et al.
and keep temperatures well below 100°C. When the waste has aged sufficiently, the chimneys are sealed and the residual thermal load is taken up by the rock. The advantage of elimination ofthe intermediate store must be weighed against the extra shafts:these have to be filled and in the long term could provide anescape route for activity.
7. CIVIL ENGINEERING
There are a number of areas in Britain where crystalline rock should prove suitable for the construction of a repository. For above-sea siting, two access tunnels each above 5 m diameter, enter the hillside for a distance of 300 m or more, leading to parallel transport galleries 800 m long and 250 m apart. These galleries will be cross-connected at 50 m intervals by 14 emplacement tunnels, each about 4 m in diameter (see Fig.3).
The emplacement tunnel arrangement for the deep repository would be similar. Access from ground level would be by twin 6 m diameter shafts with a third similar shaft providing ventilation (see Fig.l). The canisters would be disposed either directly in the 14 connecting tunnels or within vertically bored holes 20 m apart in the base of these. It might be preferred to form the vertical holes by raised boring from small headings below the tunnels. The three shafts, whilst providing more than sufficient access for ventilation and removal of spoil during the excavation of the tunnels and galleries, would also allow for complete separation of emplaced radioactive canisters from the construction of the second half of the repository. Any of these arrangements would be sufficient to accommodate the first half of the waste.
A detailed geological survey of any proposed site would be necessary - to establish amongst other things the character of the rock, its permeability at various depths, fissure pattern, modulus of elasticity and Poisson's ratio. If possible, before the detailed design and method of excavation are decided^ the magnitude of the tectonic stresses at depth should be measured.In strong, hard, crystalline rock of high compressive strength and modulus of elasticity, problems may arise because of the high in situ stresses at depths greater than 300 m. These stresses can affect both safety during tunnelling and the integrity of the completed works over the period when access and ventilation would be required. Hydraulic fracture methods may be used to estimate the magnitude of the vertical in situ stresses at the depths metioned. However the near-horizontal stresses that may be present (these have been known in certain instances to be up to ten or more times greater than the vertical) cannot be determined by means of boreholes with the same degree of confidence.
IAEA-SM-243/93 475
It is essential to avoid areas of adverse tectonic stresses and methods for determining stresses at the base of deep boreholes need to be developed. Shafts will be sunk by conventional means - making use of a multiple drill rig which will allow the precise positioning of the blast holes in any pattern. The stresses at tunnel level can then be verified in short pilot headings.
All excavations need to be carried out with minimum disturbance to surrounding rock using pre-splitting or smooth blasting methods. The back filling of galleries, tunnels and shafts will be with aggregate formed from the spoil mixed with a variety of minerals known to be effective in absorbing radioisotopes. Where necessary this will be done by remote controlled mechanical devices. It may be advantageous to grout at high pressure with a mixture of finely ground absorbing minerals and bentonite into the fill.
8. SAFETY STATUS
The safety analysis of repository concepts (after sealing)covers the following aspects:
i the structural stability of the repository rock against thermal stresses: corresponding calculations are still in progress (2) .
ii assessment of risk and consequence of low probability events. A fault tree analysis is being developed for the short term period. The survey of unlikely natural events initially set out in Ref(3)is being expanded.
iii the thermal effects of the repository on surrounding ground water is being analysed (2). Initial results suggest that thermal convection in water above the repository could provide a driving force for water movement to the surface.
iv the analysis of activity escape through ground water movement (3) is being expanded to take account ofthe above thermal convection effects and applied to Wet Repository concepts. A parallel assessment is in hand for the Dry Repository concept for the delay of activity in absorber beds and permeable strata. Preliminary results for wet and dry cases are similar in that, in the long term, the escape of isotopes such as Tc99 and Np237 cannot be discounted. The Ra226 in canisters is unlikely to build up from uranium residues to a level significantly greater than the natural Ra226 in the plane of the repository. Several actinide chains known naturally develop in the repository. However,
476 BEALE et al.
such isotopes from these that may escape add only a minute increment to the natural isotopes in sea water. The great majority of the initial radioactivity in the repositories, however, will decay away within the rock.
v Modelling the pathways to man from activity releasedto aquifers or local seas is in hand to evaluate the corresponding detriment.
9. CONCLUSIONS
Only tentative conclusions can be drawn at this interim stage of the studies.
i The working rock temperature limit of about 100°C is not exceeded at any time for any of the arrangements outlined.
ii The studies so far have not revealed any insuperable engineering problems. The assessment of the response of the rock to the repository arising from construction and thermal loading is at an early stage.
For all the concepts studied, it seems likely that adequate safety arguments can be mounted for the retention of most radioisotopes in rock. In the long term, a few isotopes, not naturally occurring, could escape, eg Tc99 and Np237. Though current evaluations may show the corresponding detriments to be small, development of good absorbers for such isotopes would provide a worthwhile safety improvement. Corresponding leakages of naturally occurring radio isotopes will provide only minute increments to existing concentrations.
iv At this stage the designs discussed are still allcandidates for the construction of a repository.
REFERENCES
[1] CHAPMAN, N.A., “ Mineralogical and geochemical constraints on maximum permissible
repository temperatures” , these Proceedings, SM-243/28.
[2] BEALE, H., et al., “Thermal aspects of radioactive waste disposal in hard rock” ,
these Proceedings, SM-243/26.
[3] H ILL, M.D., et al., Preliminary Assessment o f the Radiological Protection Aspects
of Disposal o f High-Level Waste in Geologic Formations, National Radiation Protection
Board Rep. NRPB-R69 (1978).
IAEA-SM-243/93 All
DISCUSSION
F.L.H. LAUDE: For how many years are you going to use natural convection
to cool the glass blocks? After this period how are you going to seal the chimneys?
J.R. GRIFFIN: The in situ cooling by natural convection will go on for
70 to 100 years. We have not yet designed a system for sealing the chimneys;
however, the use of a granite aggregate with grouting looks attractive.
R.G. CHARLWOOD: The range of design concepts presented, particularly
the cooling options, is very interesting. In our studies we found that the under
ground design could easily accommodate dispersion of waste, and hence lower
temperatures, at minimal extra cost. Have you made cost comparisons for your
designs to compare surface storage costs with incremental underground costs?
J.R. GRIFFIN: We have not completed the cost comparisons; we hope
to do so early next year.
A.G. JACOBI: What is the material proposed in the British concept for
the packaging of high-activity waste canisters?
I understood you to mean that the thicker packaging of high-activity waste
canisters ensures sufficient shielding and hence allows access to the end product
during transport and handling. If the thinner packaging does not permit the
repository operator to have such access, what criterion was applied for selecting
the thickness of the wrapping?
J.R. GRIFFIN: The overpack material we propose is malleable cast iron
with the possible addition of small quantities of copper.
The criterion for the thin overpack (100 mm) is that it should have a
corrosion life in the repository environment of a minimum period of 1000 years or,
in other words, that it should protect the glass against commencement of leaching
for 1000 years. Not only is the thick (250 mm) overpack adequate from a shielding
point of view to allow man access, but it also affords greater resistance to pene
tration by corrosion.
IAEA-SM-243/118
THE NUCLEAR WASTE DISPOSAL
STUDY PROJECT OF THE
GEOLOGICAL SURVEY OF FINLAND
H. NIINI
Geological Survey of Finland,
Espoo, Finland
Abstract
THE NUCLEAR WASTE DISPOSAL STUDY PROJECT OF THE GEOLOGICAL SURVEY
OF FINLAND.
With its two nuclear power plants in operation and two more units under construction,
Finland is and will be producing nuclear wastes of all levels. For the planning and control of
the waste storage and terminal disposal a national geological nuclear waste disposal study
project was started in 1977 at the Geological Survey of Finland. The project is divided into
ten sub-projects with different themes, background, time-table, and methods of realization.
Most of the sub-projects are accomplished in close co-operation with other state institutes
or the power companies, particularly with the Technical Research Centre of Finland.
Preliminary results of the most urgent sub-projects indicate that the possibilities of storing
high-active waste deep in certain rock formations in the Finnish bedrock appear relatively
good. An inventory of the existing mines and underground space reveals that certain
prospects for using abandoned mines for the disposal of low-active waste exist. The loose
Quaternary deposits of Finland may however be suitable only for the burial of very low-
active nuclear wastes.
1. BACKGROUND AND AIMS
At the present time, there are two nuclear power plants in operation in
Finland; a 420 MW pressurized-water reactor at Loviisa, SE Finland (started
in 1977), and a 660 MW boiling-water unit at Olkilouto, SW Finland (started
in 1978). At both localities a second exactly similar unit is under construction.
Thus Finland is and will be producing nuclear wastes of all levels.
The responsibility for the. planning and realization of waste disposal lies
with the power companies producing wastes [1]. So far, an agreement has been
made to send the spent fuel of the Loviisa plant to the Soviet Union. All other
radioactive wastes produced in Finland should be disposed of by Finland itself.
The most likely solution is the building of one or more waste repositories under
ground. To check the plans the government authorities need adequate basic
information concerning the possibilities of, and requirements for, the storage
and disposal of radioactive waste in the different geological formations in
Finland. This is why the Ministry of Trade and Industry in 1977 took the
479
480 NIINI
initiative of establishing a national geological nuclear waste disposal study
project, which was placed with the Geological Survey of Finland.
In addition to this project of the Geological Survey, the two power com
panies naturally have a geological programme for waste disposal. They have
already produced a general report dealing with the management of radioactive
waste from the Finnish nuclear power plants [2]. Certain plans have also been
presented for studying the use of new substances, for example ash produced by
coal power stations, as a component of filling material in connection with
radioactive waste disposal. The geological work in the programme of the power
companies is carried out in close co-operation with the project of the Geological
Survey of Finland.
2. DIVISION INTO THEMES
For 1979, the nuclear waste disposal study project of the Geological
Survey is divided into the following sub-projects, each of which is led by a
responsible geologist. The present organization of the project is shown in
Fig. 1. The sub-projects form four groups.
Group 1. General prerequisites for geological waste disposal
The evaluation of the general geological conditions from the point of view
of nuclear waste disposal will be based on the results of 7 sub-projects. In view
of the urgent needs of the power companies and the decision-making political
authorities a preliminary report on the subject was compiled in 1978 [3].
According to the preliminary estimations, the predominant hardness and
toughness of the Finnish rocks makes underground excavation sometimes
troublesome, whereas the average stability and tightness of bedrock appear
advantageous. The nuclide sorption capacity of the rocks and bedrock fractures
is not yet well known, but because of the scarcity of groundwater and its minimal
pressure gradient those radionuclides reaching groundwater will have extremely
small chances to migrate. Thus, in general, Finnish bedrock offers relatively
good prerequisites for the terminal disposal of nuclear wastes.
As to the loose Quartemary deposits, only in respect of the very low-active
wastes can the least permeable formations be considered utilizable.
Sub-project 1.0. Suitability o f rock formations and influence of
different geological and related processes
The sub-project is meant to be a summary of investigations of all geological
rock factors. It will be based on the future results of the other sub-projects
dealing with bedrock.
IAEA-SM-243/118 481
FIG .l. The organization o f the nuclear waste disposal study project o f the Geological Survey o f Finland in a simplified form. Individual scientists may be involved in a number o f these areas simultaneously.
482 NIINI
Sub-project 1.1. Tectonics and mineralogy o f joints and
shear zones in bedrock
The aim of the study is to elucidate quantitatively the natural features and
variability of the Finnish bedrock discontinuities deemed important for the use
of bedrock for storage purposes. The study is being carried out in co-operation
with the Department of Geology of the University of Helsinki (Dr. R. Uusinoka).
In fact, the study is a continuation of a preliminary study of rock brokenness
by the author [4] with more material and a more thorough statistical treatment.
The study consists of road-cut outcrops in three cross-section zones roughly
oriented through the Svecofennian schist belt in southern Finland.
At each observation site a rock body or unit showing homogeneous
jointing was determined. For each homogeneously jointed rock body the
following features were observed, numerically measured, or calculated as
systematically as possible: joint frequency and its variance, mutual orienta
tion, openness and regulatiry of the joints, rock types, degree of weathering and
its variance, and dimensions of the homogeneously jointed rock bodies. Since
observations were made both on the natural surface of the rock outcrops and
on the excavated fresh surface nearby, the total number of measured or
observed rock variables amounts to 16. The statistical treatment being made
will show the distribution and mutual connections of the variables — a piece of
information still lacking but apparently fundamental for the planning of
demanding underground excavations.
Necessary information about the joints, shear zones, and filling materials
is also being acquired from deeper rock excavations, especially from the southern
section of the Pâijânne-Helsinki water-supply tunnel being mapped by the
Petrological Department of the Geological Survey [5]. This work is connected,
and will be utilized together with the pre-construction studies on rock
fracturing [6, 7] and weathering products [8,9] at the tunnel line. This tunnel
mapping will give a continuous 120 km long profile showing detailed rock
conditions of the Finnish Precambrian at a depth of 30 to 150 metres.
Sub-project 1.2. Sorption o f radionuclides in bedrock
The sorption capacity of rocks and joint fillings is being studied with the
aid of borehole experiments and laboratory tests. Tritiated water as a tracer of
groundwater, 327Pu as a tracer of plutonium and 150Nd as a simulating carrier
of plutonium were fed into a borehole in a system of eight holes bored into a
granitoid rock on the island of Kuninkaankartano in a suburb of Helsinki. The
purpose of these field experiments is, above all, to gain experience in labelling
techniques, sample collection, sample preparation, measurement and modelling,
IAEA-SM-243/118 483
as well as of other practical details of a close co-operation between two
institutes — in this case the Geological Survey of Finland and the Department
of Radiochemistry of the University of Helsinki. Another part of the co
operation consists of laboratory tests with fresh and weathered rocks and joint
fillings. On the basis of this sub-project, more demanding site studies are being
planned for the sites considered promising for the underground disposal of
nuclear wastes.
Sub-project 1.3. The influence of land uplift and glaciation
This sub-project will be started in the summer of 1979. It will be chiefly
based on published material.
Sub-project 1.4. Flow o f groundwater in bedrock
In the spring of 1979 regular observations and measurements of ground
water in bedrock were started, the purpose of which is to elucidate the hydrological
and geological role — recharge, movements, residence time, storage, and discharge —
of groundwater in bedrock in different physiographic environments and to give
a geological background for the interpretation and applicability of the point
measurements of groundwater. At this stage only existing groundwater or rock
investigation holes and abandoned wells are being used. Naturally, the study
comprises a detailed determination of the hydrological and geological conditions
around the chosen observation points. It is partly based on observations from
the existing network of surficial groundwater stations of the National Board
of Waters. The results of the sub-project should gradually yield information
necessary for the future repository siting studies, which include deep ground
water measurements carried out in co-operation with the Technical Research
Centre of Finland in particular.
Sub-project 1.5. Disposal o f medium- and low-active wastes in
Quaternary deposits
As mentioned previously (group 1), the study is concentrating on the least
permeable deposits. At this initial stage it is mainly based on the information
on the structure and properties of Finnish soils already published or otherwise
available. A limited field study programme will be started in the summer of
1979. In addition, sampling of certain clay formations as well as some
laboratory determinations of the Kd-values of the clay samples are planned.
484 NIINI
Sub-project 1.6. Regulation o f groundwater flow by
excavation and drilling
This sub-project comprises simplified model experiments aimed at eluci
dating the idea that the groundwater flow in the repository rock could be
artificially minimized by boring a network of holes around the repository so
that if groundwater movements occur, their effect would concentrate on these
holes and leave the repository itself untouched. It is thought that such a system
would operate without maintenance. In our experiments we intend to substitute
sand or some porous material for rock and use pipes of porous material as drill
holes so that their density and configuration can easily be changed in the
experiments. The study is meant to be carried out in co-operation with the
Laboratory of Hydrology and Water Management of the Helsinki University of .
Technology.
Group 2. Regional surveys
Sub-project 2.1. Inventory o f existing rock openings
(mines and other underground space)
The sub-project aims at evaluating the possibilities of using existing mines
or corresponding underground facilities for temporary storage or terminal
disposal of the wastes. A preliminary report has been produced [10]. As to
the high-active wastes, the results were fairly negative, whereas in the case of
the medium-active and low-active wastes the inventory is still insufficient and
is being continued in close co-operation with the Technical Research Centre of
Finland and the power companies. The criteria influencing the suitability of
a mine as a waste storage site have been classified as follows:
(1) The excavation and construction operations carried out, and their
results: rock openings and stopes, their dimensions, form, depth, and
technical construction.
(2) The natural bedrock: rock types, geological structures, groundwater
and strength properties.
(3) Environmental factors: topography, hydrography, geographical
position, human activities and conditions in the area.
At the next stage of the study only sites situated in southern Finland —
reasonably close to the existing nuclear power plants — will be investigated. A
notable part of the study is formed by the determination of the evaluation
criteria for different kinds of wastes in both abandoned, operating, and future
underground excavations. These criteria are being determined by the organiza
tions mentioned, together with the Institute of Radiation Protection.
IAEA-SM-243/118
Group 3. Local investigations
485
Plans have been made to start local geological investigations at the two
sites of the nuclear power plants. Naturally, they will be performed in
co-operation with the power companies. This does not suggest that any
decisions on the use of these localities for the terminal disposal of nuclear
wastes have been made, but it is in accordance with the idea that it would be an
advantage if the final repository could be located as near a power plant as
possible. Accordingly, the local investigations planned so far will comprise a
field survey at both Olkiluoto and Loviisa. These sub-projects will be defined
later.
Group 4. Development work and international co-operation
A very important part of the work within the project is co-operation with
international bodies such as the International Atomic Energy Agency, OECD
Nuclear Energy Agency, and the International Association of Engineering
Geology (IAEG). Essential material has been acquired from those countries
that have roughly similar geological conditions to Finland, such as Sweden,
Canada, and to some extent the USA. One sub-project defined in greater detail
is in operation:
Sub-project 4.1. Development o f suitable geological investigation
methods
The sub-project is partly based on close co-operation with corresponding
Swedish authorities. As to the geophysical and technical equipment, the sub-
project is connected with a similar one of the Technical Research Centre of
Finland.
3. CONCLUSION
The organization, economic resources, and future plans for the nuclear
waste disposal study project of the Geological Survey of Finland are by no
means strictly fixed, but rather, the project is at an initial, dynamic stage
subject to changes. The project directly follows the instructions of the Energy
Department of the Ministry of Trade and Industry, which in late 1978
established a Co-ordinating Group to link together all the nuclear waste study
projects carried out by different institutes and companies in Finland and to
make a collective long-term plan. It is anticipated that the geological study
project with the aims described will be in operation for approximately one
decade at least.
486 NIINI
The results of the study concerning the sub-project themes or parts of them
are intended to be published preferably in concise reports as soon as appropriate
parts of the study are finished.
ACKNOWLEDGEMENTS
The author is indebted to many of his Swedish colleagues — especially
Drs. Otto Brotzen and Ulf Thoregren — for necessary guidance and support at
the initial stage of the project, and to Professor L.K. Kauranne, Mr. Martti Salmi,
and Miss Anne Paksuniemi for helpful remarks concerning the text.
REFERENCES
[ 1 ] Arrangements of Nuclear Waste Management in Finland, Ministry of Trade and Industry,
Energy Dept., Series B:4, Helsinki (1978) 1—57.
[2] Management of Radioactive Waste from Finnish Nuclear Power Plants, Rep., Imatran
Voima Oy (IVO Power Company) and Teollisuuden Voima Oy (TVO Power Company),
Helsinki (1978) 1-227.
[3] N IIN I, H., Suomen geologiset olosuhteet ydinjâtteiden varastoinnin kannalta (Geological
Conditions in Finland from the Point of View of the Disposal of Nuclear Wastes),
Geological Survey of Finland, Duplicated Rep. (1978) 1—21.
[4] N IIN I, H., “Engineering-geological classification and measurement of the brokenness
of bedrock in Finland” , Engineering Properties and Classification of Natural Materials
of Construction (Proc. 2nd Int. Conf. Sao Paulo, 1974), Vol. 1, Theme IV, Int. Assn.
Eng, Geol. Sao Paulo (1974).
[5] SUOMINEN, V., POKKI, E., Geological Observations from the Paijanne Tunnel,
Second Building Stage, Geological Survey of Finland, Rep. of Investigation (1979)
(in preparation).
[6] N IINI, H., “On engineering-geological studies concerning the selection of the water
tunnel Hausjârvi-Helsinki” , Engineering Geology, Amsterdam, 2 1 (1967) 39—45.
[7] N IINI, H., “A study of rock fracturing in valleys of Precambrian bedrock” , Fennia 97 6
(1968) 1-60, Eng.-Geol. Soc. Finland 3 26 (1968).
[8] UUSINOKA, R.-, “A study of the composition o f rock-gouge in fractures of Finnish
Precambrian bedrock” , Commentationes Phys.-Math., Helsinki 45 1 (1975) 1 — 101.
[9] N IIN I, H., UUSINOKA, R., “Weathering of Precambrian rocks in Finland” , Properties
of Soils, Rocks and Rock Masses (Proc. 3rd Int. Conf. Madrid, 1978) V o l.l, Sec.II,
Int. Assn. Eng. Geol. Madrid (1978) 77-83.
[ 10] N IIN I, H., Suomen kaivostilojen inventointi ydinjâtteiden varastoinnin kannalta (An
Inventory of the Finnish Mines Considering Their Possible Use for the Disposal of
Nuclear Wastes), Geological Survey of Finland, Duplicated Rep. (1978) 1 — 20.
IAEA-SM-243/22
BENTONITE-BASED BUFFER SUBSTANCES FOR
ISOLATING RADIOACTIVE WASTE PRODUCTS
AT GREAT DEPTHS IN ROCK
R. PUSCH, A. JACOBSSON
University of Luleá,
Luleá,
A. BERGSTROM
Nuclear Fuel Safety Project,
Stockholm,
Sweden
Abstract
BENTONITE-BASED BUFFER SUBSTANCES FOR ISOLATING RADIOACTIVE WASTE
PRODUCTS AT GREAT DEPTHS IN ROCK.
Bentonite is a vital sealing component of buffer masses used in the Swedish KBS project
for deposition of highly radioactive waste products. Metal canisters which contain the wastes
are placed in deposition holes bored from tunnel floors at 500 m depth in crystalline rock. In
shafts and tunnels fairly dry bentonite/quartz mixtures will be applied by means of ordinary
field compaction techniques. A bentonite/quartz ratio of 1:10 to 1:5 provides a chemically
stable isolation with the required stiffness, plasticity and permeability. In the case of vitrified
nuclear wastes bentonite/quartz mixtures are also used in the deposition holes, while highly
compacted bentonite will be used in these holes to separate the copper-mantled, unreprocessed
nuclear fuel wastes from the rock. The bentonite will swell by taking up groundwater, thus
creating a sealing substance which is practically impervious. Due to its high swelling potential it is
‘self-sealing’ in case of minor rock displacements. Joints extending from the deposition holes
will be partly filled by extruded bentonite. An additional way of using bentonite as a sealing
component is offered by the electrical charge of the montmorillonite particles of the bentonite.
Exploratory tests have indicated that bentonite can be used for sealing narrow joints by applying
electrophoresis.
INTRODUCTION
The Swedish KBS concepts for final deposition of nuclear wastes from reactors are based on the presumption that metal canisters with the wastes are placed in deposition holes bored from the floors of tunnels at about 500 m depth in Swedish crystalline bedrock (1, 2). The canisters will be isolated from the rock by a "buffer mass" consisting of bentonite or mixtures of bentonite and quartz particles (Figs 1 and 2). The deposition plant will finally be sealed by filling shafts and tunnels
487
488 PUSCH et al.
1.5
II
FIG .l. Tunnel and deposition hole with
canister for vitrified reprocessed reactor
wastes. I = 80-90% quartz (sand¡siltj
and 10-20% bentonite (clay). II = 85%
quartz and 15% bentonite. I l l :
quartz and 10% bentonite.
FIG.2. Tunnel and deposition hole with
canister for unreprocessed wastes.
I cf. Fig. 1. 11= Highly compacted
bentonite.
IAEA-SM-243/22 489
with bentonite/quartz mixtures. Bentonite is the term for mont- morillonite-rich clay formed by devitrification of the natural glass component of volcanic ash deposited in prehistoric lakes and estuaries.
The selection of suitable buffer mass compositions for the deposition holes was based on the following criteria:
ф The buffer mass must provide a protection of the canisters against falling rock fragments. It must also provide sufficient support for the canisters.
Requirement: Sufficient stiffness (high "passive earth pressure"; high bearing capacity; negligible settlement).
# The buffer mass must be sufficiently soft so that only moderate stress is produced in the canisters in case of small displacements in the bedrock.
Requirement: Sufficient softness ("plastic consistency").
# The buffer mass must have a very low permeability and dif- fusivity to minimize the ground water percolation, which determines the rate of canister corrosion, and to cut down the diffusion rate of radiotoxic nuclides which will escape when corrosion has finally exposed the canister content.
Requi rement: Non-brittle behaviour (no fissuring or crack- ing); low permeability and diffusivity; swelling ability to guarantee complete filling of the holes, especially in connection with possible, small rock displacements.
# The buffer mass should have a certain cation exchange capacity.
Requirement: A certain ability to adsorb radionuclides is valuable.
# The buffer mass should have a fairly high heat-conducting capacity.
Requirement: The heat conductivity must be sufficiently high so that the elevated temperature produced by the canister does not substantially change the properties and intended functions of the buffer material.
There are considerable difficulties in handling and applying large quantities of buffer material in tunnels and shafts.
490 PUSCH et al.
A А А Л
LEG EN D
@ О Н
О О
° M g , A l
• S I A l
FIG.3. Probable crystal structures of montmorillonite, cf. Ref.(3). Infrastructural space contains
water molecules and cations.
Since these masses are only slightly influenced by radiation and elevated temperatures, a lower degree of homogeneity and permeability must and can be accepted than for the buffer material in the deposition holes. In principle, however, they have similar functions.
I.t was obvious at an early stage of the investigation that a stiff, homogeneous montmorillonite-rich clay, such as bentonite, would be a suitable buffer material, provided that it retains its physical and mechanical properties for extremely long periods of time (lO^-lO^ years or more) despite the radiation and the elevated temperature. The main question concerned the consistency required to obtain a bed of stiff clay with a high degree of homogeneity. A suitable bearing capacity and suitable stress/strain properties correspond to a water content of no more than 50-100% if water saturated Na bentonite is used, but in situ compaction of such a moist clay cannot be performed with the required high degree of homogeneity. This difficulty can be eliminated by using fairly dry clay powder. If such powder is applied in layers which are compacted, a dense and homogeneous mass is obtained which then slowly absorbs water from the confining rock without losing its homogeneous character.
IAEA-SM-243/22 491
The ability of the clay mineral montmorilIonite to take up water and swell (cf. Fig. 3), which is especially strong when Na is the main adsorbed cation, requires some sealing arrangement to keep the total volume constant,since the necessary low water content would otherwise be largely exceeded. This can be arranged for the deposition holes but hardly for tunnels and shafts. The density of in situ compacted "dry" clay in the deposition holes would, however, be toolow to yield a sufficiently high heat conductivity. The temperature would exceed 100 C, which is a safe value to avoid any change of the tetrahedral and octahedral members of the crystal structure (4).
BENTONITE/QUARTZ AS BUFFER SUBSTANCE
The obvious difficulty in using only bentonite suggested the introduction of an additional mineral component. Quartz was considered to be suitable since it is a silicate with very good heat-conducting properties and a considerable chemical stability. It was concluded, however, that the quartz part of a bentonite/quartz mixture had to be dominant to yield the required heat conductivity and this would reduce the beneficial properties of the bentonite: the low permeability and the high ion exchange and adsorption capacities. On the other hand, a suitable, till-like quartz particle grading would give a high bulk density and at least a fairly low permeability. This principle was finally chosen for the composition of all the buffer materials in the KBS concept for vitrified, reprocessed nuclear wastes (Fig. 1) and for the tunnels and shafts in the KBS concept for unreprocessed fuel wastes.
A comprehensive literature survey and a number of tests of bentonite/quartz mixtures with a weight ratio varying from approximately 1:10 to about 1:5 (10-20% Na bentonite and 80-90% silt/sand-sized quartz particles) were made to investigate the influence of pH and radiation on the stability of the two minerals. It was found to be quite satisfactory for pH values ranging between 7 and 9, which are reasonable limits for the ground water in Swedish crystalline bedrock. Experiments with high gamma radiation doses gave only insignificant effects on the X-ray diffraction patterns and no aluminium traces indicating lattice destruction have been found (5).
The plastic properties of the buffer substances must be preserved to prevent the formation of cracks and fissures. Such permeable passages can easily be produced even by minor rock displacements if the buffer substance has turned into a brittle condition. A transition of this kind can be produced by cementation either by the action of high grain contact pressures ("pressure solution") or by solution/precipitation of SiO2 from
492 PUSCH et al.
closely situated quartz grains. Only the latter process can be of any importance since the grain pressures will be very moderate. It was found, however, that the extent of SiC^ solution with secondary precipitation in quartz grain contacts will not be sufficient, even after a million years, to result in brittle behaviour of the bentonite/quartz mixtures, despite the markedly increased solubility of quartz which takes place if both pH and temperature are raised (6).
The conclusion that the long term chemical stability can be safely predicted is strongly supported by the well-known occurrence of natural plastic clay seams with montmorillonite and quartz as dominant constituents that are several million years old. Such seams, which result from hydrothermally effected weathering, are frequently observed in granite and gneiss in Sweden.
The most important physical properties of bentonite/quartz mixtures, such as the permeability, shear strength, compaction properties and swelling pressures, are fairly well-known from the literature. Additional information concerning the particularly interesting permeability, diffusivity and swelling properties was obtained by performing various investigations.These were made by using a commercial bentonite (granulated Volclay MX-80 produced by the American Colloid Co) with sodium as the main adsorbed ion. The grain size distribution of the quartz component used in the experiment and suggested for practical use is illustrated by Fig. 4. The water content of the mixtures averaged 5-20% while the dry density ranged between 1.4 and 1.7 t/щЗ, the latter corresponding to what can probably be achieved by applying field compaction techniques in the tunnels and shafts.
It was found that the permeability of completely or almost water-saturated bentonite/quartz mixtures with a weight ratio varying from 1:10 to 1:5 ranges between 10"^ to 10“'’ m/s, depending on the density and the hydraulic gradient (4). Such buffer materials do not obey Darcy's law, which means that the extremely low hydraulic gradients (10-2 to 10"3 ) which will exist some hundred years after the closing of a deposition plant, will hardly produce any water percolation at all. Partial or complete ion exchange from sodium to calcium caused by a largely changed ground water chemistry, or an exchange to nuclides from corroded canisters will produce a certain but not very important increase in the permeability. This is also the case if the electrolyte concentration is increased as a consequence of largely changed climatic or topographical conditions by which the area maybe flooded by sea water. A separate study showed that there is no risk of substantial clay particle transportation and piping caused by temporary high hydraulic gradients which may exist after the closing of the deposition plant and in course of the filling and compacting operations (7). This study
IAEA-SM-243/22 493
0.002 0.02 0.2 2 20
GRAIN SIZE, mm
FIG.4. Grain size distribution for the quartz component of bentonite ¡quartz buffer masses.
also illustrated the significant self-sealing property of the 1:5 bentonite/quartz mixture in case of local piping.
The insignificant water percolation, which results from the very low permeability coefficient, means that diffusion is the decisive ion transport mechanism. The diffusivity of bentonite/ quartz mixtures with a weight ratio ranging approximately from 1:10 to 1:5 was found to be about 1/10 of that of water for metal cations (8). The swelling properties, expressed in terms of swelling pressure produced by a confined sample which is exposed to water, was found to be very moderate. For the 1:10 mixture it does not exceed a few hundred kilopascals, while it will be several hundred kilopascals for the 1:5 mixture.These values are preliminary only.
HIGHLY COMPACTED Na BENTONITE AS BUFFER SUBSTANCE
Unreprocessed nuclear fuel wastes must be more effectively
isolated from the biosphere. This can be achieved by improving the "engineering barrier", which again suggests the use of pure Na bentonite. The previously mentioned difficulty concerning the heat conductance can be eliminated if the bulk density and degree of water saturation are largely increased. The Swedish company ASEA-Atom suggested and produced blocks of highly compacted Na bentonite and a number of such specimens have later
494 PUSCH et al.
been made and tested by others. Pressures in the range of 50-100 MPa gave densities of the order of 2.0-2.3 t/m^ when using air- dry (10% water content) MX-80 powder. The blocks, which have a fairly high degree of water saturation, seem to remain stable and do not dry out or take up additional water when the relativehumidity of the atmosphere is 40-60%.
The KBS concept (cf. Fig. 2) implies a block bulk density of 2.2 t/nr at 10% water content when using MX-80. The degree of water saturation, which is 60-70% at this stage, will finallyincrease to 100% when the deposition plant has been closed fora sufficiently long time and high water pressures have been established.
The particularly attractive properties of the highly compacted Na bentonite are:
# Extremely low permeability
# Low diffusivity
ф Considerable ion adsorption capacity
# High swelling and self-sealing capacities
There is considerable information in the literature concerning the permeability of water-saturated Na bentonite of various densities. Fig. 5, which presents the result of a literature survey, indicates that this clay type, for which Darcy's law is not valid, is practically impervious for bulk densities higher than about 2 t/nr. As in the case of bentonite/ quartz mixtures, the diffusivity is therefore the only ion transport mechanism of interest. The diffusivity has been found to be about 1/100 of that of water for metal cations (8). This is a considerable improvement as compared with bentonite/quartz mixtures and it makes the highly compacted bentonite barrier very effective, especially considering the high ion exchange capacity, 80-100 mea/100 g. Thus, following (8), the nuclides Sr^O, Cs 137 and /\m241 w -¡n be retarded so much that they have
practically decayed before escaping from the bentonite barrier. The transport of R a ^ o and Pu^^O is also considerably retarded. The very low permeability and diffusivity of dense bentonite is explained by the extremely small interlamellar spacing. Thus, a density of 2 t/vfi of water-saturated Na bentonite means that the water content is 25%, which corresponds to an average inter- lamellar 3-5 A water film only, because of the very large specific surface area of montmorilIonite. The water molecules constituting this film will be strongly adsorbed to the clay mineral surfaces, which leaves only narrow, tortuous interparticle passages for water and ion transportation (Fig. 6).
IAEA-SM-243/22 495
k, m/s
FIG.5. Average relationship between coefficient of permeability (k) and bulk density (p) for
Na bentonite. The hatched area accounts for scattering and variation in testing technique.
FIG.6. Electron micrograph of ultra-thin section of clay with a high amount of Na
montmorillonite and with a bulk density of about 2 t/m3 (water-saturated). Dark objects
represent clay particles, each consisting of a large number of laminae shown in Fig.3.
496 PUSCH et aL
FIG. 7. X-ray picture of model test simulating the processes which take place as a consequence
of an instant displacement which produces local openings (A) when the canister moves in a
buffer mass. I f it consists of highly compacted bentonite, self-sealing will finally lead to a
homogeneous condition. White dots are caused by lead shots for displacement study. Length
of model canister 10 cm.
The highly compacted bentonite will take up water from the surrounding rock with tremendous power. This is accompanied by strong swelling which produces a pressure on the canisters and the confining rock as well as on the overlying tunnel fill with the same composition and properties as in the concept for vitrified, reprocessed nuclear wastes. The water uptake increases the bulk density of the compacted bentonite in the deposition holes to about 2.3 t/m3 but since the swelling pressure will displace the tunnel fill a few decimeters, the corresponding swelling will make it drop to about 2.1 t/m3 in the final condition of equilibrium (9).
The practical consequences of the swelling produced by the water uptake are of great importance. Firstly, swelling makes the bentonite fill the deposition holes completely,also if minor rock displacements should occur. A temporary, local reduction of the density or even formation of a local open space may be caused by such a displacement but the self-sealing property guarantees that such defects disappear and that uniform conditions are finally established (Fig. 7). Secondly, swelling bentonite enters and seals rock joints which may be opened by displacements or changed stress conditions in the rock. Provided that the suction potential is constant, such an extrusion of bentonite will take place irrespectively of whether the joint widening occurs during the deposition or 100 000 years later.
IAEA-SM-243/22 497
30
20(QQ.s
Q t O 1 0
01. 5
FIG.8. Observed swelling pressure (ps) for highly compacted Na bentonite. The bulk density p
refers to the condition o f complete water saturation.
The rate of extrusion into rock joints is known to be a function of their width (10). For narrow joints the dominant physical process is that of viscous flow with the swelling pressure gradient as the main driving source, while for joints wider than a few millimeters, simple expansion by water uptake
governs the rate of extrusion, which will probably not proceed to more than a few decimeters depth. Yet, it is known to contribute largely to the low permeability and diffusivity of the shallow parts of the confining rock. The limited depth to which bentonite will be extruded even when the joint width is considerable means that no large bentonite loss will take place.
One of the most important parameters in the future design of deposition plants is the swelling pressure. It is known to depend on the montmorillonite content, bulk density, particle orientation and type and amount of adsorbed cations. Fig. 8 shows bulk density versus swelling pressure for highly compacted blocks of Na bentonite (MX-80). The correlation is approximate only, since the testing technique largely determines the relationship. The pressure can be roughly estimated by applying electrical double layer theories. It increases with increasing bulk density and degree of particle orientation and is much higher for monovalent than for di- and polyvalent adsorbed cations. Current research in the Division of Soil Mechanics, University of Luleâ, is intended to improve the accuracy of such theoretical predictions.
The bentonite in the deposition holes will exert a pressure of 10 MPa at maximum. In the final condition of equilibrium, which is reached after swelling produced by the displacement of the tunnel fill, the swelling pressure will probably drop to about 5 MPa. This moderate pressure eliminates the risk of tangential tension stresses in the rock at the periphery of
0 1.5 2.0 2.
p , t / m 3
498 PUSCH et al.
the deposition holes, provided that the primary stress condition in the rock is fairly isotropic.
A uniform swelling pressure on the confining rock and on the canister will be produced since the bentonite blocks will be separated from the rock and canister by narrow slots filled by Na bentonite powder which is more permeable than the dense blocks. Thus, water will first be taken up and distributed in these slots by which uniform water access is provided for the subsequent water uptake in the blocks.
Chemical stability is of course required for the highly compacted bentonite as well as for the bentonite/quartz mixture. Radiation, pH and elevated temperature will affect both buffer materials similarly. Electrolyte changes, on the other hand, may have a stronger influence on the highly compacted bentonite than on the mixture, especially with respect to the swelling pressure. Thus, it will probably be substantially lower if Na is completely replaced by Ca. The negative effect of ion exchange on the physical properties of the highly compacted bentonite is minor, however, and can easily be taken into consideration. The impression is therefore that the isolating power of highly compacted Na bentonite is extraordinarily good. Its efficiency may even turn out to make the retaining function of the huge rock mass superfluous. The main role of the rock would then only be to provide a safe mechanical protection for the "chemical apparatus".
SEALING OF ROCK JOINTS
The isolating power will be largely improved if joints extending from the deposition holes can be sealed effectively.This could be achieved by filling the joints with a gel within 1-2 m from the hole periphery. Even a moderate density of such a gel would reduce the average diffusivity and permeability to such low values that there will be practically no water or ion transport to or from the confining rock.
The problem is to seal joints with a width smaller than 0.01-0.1 mm. Bentonite suspensions cannot usually be effectively injected by applying pressure in such narrow joints. One possible technique, which is based on electrophoresis, was suggested by one of the authors in the survey of rock stabilization methods
initiated by the KBS project (11). A pilot test with glass plates to simulate rock joints, and a small scale experiment on a 1 nr block of diorite were made. It was found that montmorilIonite particles could be effectively transported from a bore hole with a dilute bentonite slurry out into joints extending from this hole. The transport of the negatively charged clay platelets
IAEA-SM-243/22 499
T IM E , h
FIG.9. “Falling head" permeability test o f rock before and after electrophoretic bentonite
sealing. Draw-down time before treatment is about 12 min and after treatment about 9 hours.
was due to an electrical potential field produced by means of electrodes placed in the hole with the slurry (cathode) and in a circumferential row of bore holes (anodes). The electrophoretic treatment largely reduced the permeability of the joints as shown by Fig. 9. A large-scale field test is presently being prepared in Sweden.
Long-term stability of the injected or extruded bentonite gels in rock joints is essential for maintaining low permeability and diffusivity. Theoretical considerations and experiments indicate that bentonite which expands freely when exposed to water will probably retain a gel character although the water content may be largely increased by the expansion. Sodium saturated montmorilIonite, for instance, is known to flocculate in ground water even when the clay concentration is reduced to a fraction of 1 per cent.
THE RAW PRODUCT BENTONITE
The experimental KBS studies were made by using Volclay MX-80, which is obtained by applying simple refining techniques to bentonite obtained from natural deposits in Wyoming and South Dakota in the USA. Other types of similar bentonites can probably be used as well, which means that the raw product resources are very good. The world production in the period 1973-1975 averaged 4.2 million tons per year.
MX-80 is a "granulated" Na bentonite, 80% of which consists of montmori1 Ionite with Na as dominant, adsorbed cation.
500 PUSCH et al.
Its chemical composition according to the manufacturer's analyses is as follows:
Component
SÍO2
A12°3
pe2°3
FeO
Ti02
CaO
MgO
Na20
k 2o
Crystal water
Additional
Perccent
~ 63
~ 21
3.2
0.3
0.1
0.7
2.7
2.2
0.4
5.6
0.8
Hydrometer analyses after ultra-sonic treatment have shown that MX-80 has a clay content of 85-90%. Silt forms the major part of the remainder which also contains a small amount of sand
REFERENCES
[1] KBS (Karnbranslesakerhet). Kârnbranslecykelns Slutsteg. Forglasat Avfall frân Upparbetning, III Anlaggningar (1977)
[2] KBS (Karnb'ranslesakerhet). Kârnbranslecykelns Slutsteg. Slutforvaring av Anvant Karnbransle, II Teknisk del (1978).
[3] JACOBSSON, A., Clay-Water Interactions. Diss. University of Stockholm (1974).
[4] JACOBSSON, A. & PUSCH, R., Deponering av Hogaktivt Avfall i Borrhâl med Buffertsubstans. KBS Teknisk Rapport No. 03 (1977).
[5] ALLARD, B., KIPATSI, H., & RYDBERG, J., Radiolys av Utfyll- nadsmaterial. KBS Teknisk Rapport No. 56 (1977).
[6] PUSCH, R., Influence of Cementation on the Deformation Properties of Bentonite/Quartz Buffer Substance. KBS Teknisk Rapport No. 14 (1977).
IAEA-SM-243/22 501
[7] PUSCH, R., Clay Particle Redistribution and Piping Phenomena in Bentonite/Quartz Buffer Material Due to High Hydraulic Gradients. SKBF/KBS Teknisk Rapport No. 79-01 (1979).
[8] NERETNIEKS, I., Transport of Oxidants and Radionuclidesthrough a Clay Barrier. KBS Teknisk Rapport No. 79 (1978).
[91 PUSCH, R., Highly Compacted Na Bentonite as Buffer Substance. KBS Teknisk Rapport No. 74 (1978).
[10] PUSCH, R., Water Uptake and Swelling of Montmorillonitic Clay Seams in Rock. Proc. 4th Int. Congr. Rock Mech., Montreux 1979 (in press).
[11 ] PUSCH, R., Rock Sealing with Bentonite by means of Electrophoresis. Bull. Engng. Geol. (in press).
DISCUSSION
R.E. GRIM: The authors are to be commended for selecting a bentonite
with the best properties for the particular purpose. It must be remembered that
not all bentonites have the same properties. Even sodium bentonites do not all
have the same properties. Not just any bentonite selected will necessarily give the
results shown in the paper. The authors are aware of this, but the fact should be
emphasized.
The authors have made an excellent beginning in assessing the use of clays
as barriers. I hope they will continue their study with other bentonites, for
example, those with higher cation-exchange capacity and different swelling
properties, and also possibly with other clay minerals, for instance, those with
specific properties of fixing certain cations.
R. PUSCH: Professor Grim is perfectly right. I can inform him that a
systematic survey of a number of bentonites and the specification of the required
physical and mineralogical properties is in fact under way.
D.L. RANÇON: Preliminary laboratory tests ahve shown that the heat
released by the repository will make changes for better or for worse in the
properties of clays, especially the adsorbing properties. Have you any data on
such changes in the case of montmorillonite?
R. PUSCH: We know that for a temperature increase to 300—400°C the
MX-80 (Na) bentonite with 10% water content has not undergone noticeable
changes in its cation exchange capacity. We intend to make large-scale field tests
on various buffer mass functions in the Stripa mine, and these will include high-
temperature tests which, together with systematic laboratory studies, will give us
the information. In the KBS studies 100°C was taken as an upper safe tempera
ture level.
502 PUSCH et al.
H. KRAUSE: The bentonite you are using as an additional barrier contains
about 10% water. As this material will be exposed to high irradiation doses, the
water will undergo radiolysis. Have you taken this into consideration and if so,
don’t you think there is a risk of critical concentration of hydrogen in the air of
the storage area?
R. PUSCH: It should be borne in mind, first of all, that part of the original
water is strongly fixed in the minerals and, in principle, forms part of the crystal
lattices. After deposition additional water corresponding to a total water content
of about 20% is taken up. Radiation is very weak because of the lead, copper or
titanium components of the metal canisters and so the hydrogen yield is very low.
This hydrogen is dissolved in the pore water. Perhaps Mr. Ahlstrom would wish
to add to these remarks.
P.-E. AHLSTROM: The problem of radiolysis has been circumvented in
the KBS proposals by having sufficiently thick canisters. With the highly compacted
bentonite we use a 20 cm thick copper canister which will reduce the radiation
field to a negligible level. This means that the water in the highly compacted
sodium bentonite will not be subjected to any significant radiolysis. More recent
experiments suggest that there is very low net hydrogen production even in the
case where the canisters have no shielding and that this would not be a safety
problem.
W.R. BURTON: Does the thin layer of bentonite surrounding the canisters
add a significant extra degree of safety in relation to the large rock cover?
R. PUSCH: The efficiency of the bentonite with special reference to
diffusion and other transport mechanisms are described in detail in
Professor Neretniek’s paper (IAEA-SM-243/108). It is beyond doubt that even
this fairly thin bentonite shield is extremely effective.
IAEA-SM-243/166
CORROSION STUDIES ON COPPER AND
TITANIUM-LEAD CANISTERS
FOR NUCLEAR WASTE DISPOSAL
L.B. EKBOM
National Defence Research Institute,
Stockholm
K. HANNERZ
ASEA-ATOM,
Vasteras
K.S. HENRIKSON
Studsvik Energiteknik AB,
Nykôping,
Sweden
Abstract
CORROSION STUDIES ON COPPER AND TITANIUM-LEAD CANISTERS FOR
NUCLEAR WASTE DISPOSAL.
The Nuclear Fuel Safety Project (KBS) has proposed that spent non-processed nuclear
fuel shall be disposed of by enclosing it in copper canisters or alternatively that reprocessed
and vitrified waste shall be enclosed in a titanium canister with a lead lining. The canisters
are to be placed in vertical drill-holes in rock, 500 m below ground and embedded in a buffer
of sand and bentonite. The purpose of this arrangement is to raise several obstacles against
fission products reaching the biosphere. The thick-walled canister is one of these obstacles,
which is proposed to be a barrier for a considerable period of time. Corrosion is the limiting
factor of the canister durability. The rate of corrosion is dependent on the amount and
transport of corrosion reactants to the surface of the canister. The thermodynamic possibilities
for various corrosion reactions on copper and lead under the prevailing conditions were
studied, also with regard to bacterial influence. Entrapped atmospheric oxygen and sulphide
in the ground water were found to be reactants of importance. The supply of oxygen and
sulphide by diffusion was calculated, and hence the greatest possible corrosion. The corrosion
attack may start as pitting but will penetrate into the thick metal wall at a decreasing rate.
An expert group arrived at the conclusion that under given conditions the canisters will last
for a very long time (hundreds of thousands of years for copper canisters). In order to verify
the expected high corrosion resistance of titanium, laboratory tests have been carried out in
environments, which must be considered to provide accelerated rather than simulated tests.
The duration of the corrosion testing was 600 days and comprised evaluation of general
corrosion or oxidation, localized corrosion and hydrogen pick-up. The results are that no
localized corrosion or hydrogen pick-up was detected and that the oxidation rate was extremely
low, corresponding to a wall-thinning of titanium of only 0.01 /im/year. In addition, the
results of potential measurements during 7-radiation are reported. They show that the passivity
of titanium is not affected by the radiation.
503
504 EKBOM et al.
1. INTRODUCTION
The Swedish Nuclear Fuel Safety Project (KBS) has proposed two different
canisters for the disposal of nuclear waste in deep granitic rock. For reprocessed
and vitrified waste a titanium-lead canister was chosen while for the direct
deposition of nuclear fuel a copper canister was proposed [1, 2].
The canisters are to be placed in vertical drill-holes in rock, 500 m below
ground and embedded in a buffer of bentonite or sand-bentonite (Fig. 1). The
purpose of this arrangement is to provide several barriers against fission products
reaching the biosphere. The canister is one of these barriers, and the intention
has been to make it long-lasting, in contrast to most other proposals for geologic
waste isolation. Corrosion is the limiting factor in canister durability. Because
of the tectonic stability of the rock in Fennoscandian cratonic shield, mechanical
failure due to shear movement can be excluded if existing crack zones are avoided.
The Swedish Corrosion Institute has been assigned the task of estimating
the life of the canisters from the corrosion point of view. To do so, the Corrosion
Institute has appointed an expert group of nine specialists1 [3, 4].
The materials of the canisters were chosen following several suggestions.
The thermodynamic stability of copper in pure water was considered to be of
fundamental importance. To minimize the radiolysis outside the canister the
wall thickness has to be sufficient. The wall thickness is also important if
penetration of the canister due to pitting corrosion is to be avoided.
Titanium was considered a useful material because of its passivity in a wide
pH range (4—10) in oxygen-free as well as oxygen-saturated water (Fig.2). The
ability to withstand local corrosion in chloride solutions is also well documented.
Lead was initially intended mainly as a radiation shield but has been shown
to be a considerable barrier. In fact the corrosion of lead was evaluated in a
similar way to copper.
This paper will give a review of the way in which the corrosion rates of the
different canister materials were calculated as well as giving details of laboratory
tests to determine the corrosion rate of titanium in a simulated groundwater.
2. EVALUATION OF COPPER CANISTER CORROSION
The spent nuclear fuel, in the form of fuel rods from the reactor, is to be
enclosed in copper canisters. The copper canisters will be made of “Oxygen Free
High Conductivity” (OFHC) copper with a wall thickness of 200 mm. The
canisters, one in each hole, will be embedded in isostatically compacted bentonite.
1 Dr. R. Carlsson, Dr. L. Ekbom, Dr. G. Eklund, Prof. I. Grenthe, Dr. R. Hallberg,
Mr. S. Henriksson, Prof. E. Mattsson (chairman), Prof. N.G. Vannerberg, Prof. G. Wranglén.
IAEA-SM-243/166 505
V itrified waste
Stainless steel
12 m
. I f i Buffer material
Compacted buffe r material
Fuel rods embedded in lead
FIG.l. Schematic diagram o f the proposed final repository 500 m below ground in
crystalline bedrock. A number of barriers prevent and retard the dispersal of radioactive
elements from the waste.
506 EKBOM et al.
FIG.2. Potential-pHdiagram for titanium.
The tunnels will be filled with a mixture of 90—80% quartz sand and 10—20%
bentonite. Due to an oxidizing treatment the sulphide of the bentonite and
organic matter will be less than 200 mg/kg. The buffer material of the tunnels
will have an admixture of iron(II) phosphate for removal of atmospheric
oxygen initially present (Fig. 1).
When the repository is closed, groundwater from the surrounding rock
enters and saturates the buffer in the course of hundreds of years. The entrapped
atmospheric oxygen then dissolves in the water and the bentonite swells so that
cracks and pores are filled. The permeability of the buffer becomes extremely
low (less than 10“12 m/s) and mass transport through the buffer will take place
practically only by diffusion.
The groundwater will, except for an initial period, be in equilibrium with the
rock and have an oxygen content of less than 0.1 mg/1 and a low sulphide content,
at the most 5 mg/1 and the pH value in the range 7—9, Table I. The maximum
temperature of the canister will be 80°C after deposition. After 1000 years
the temperature will have decreased to 50°C.
IAEA-SM-243/166 507
TABLE I. PROBABLE COMPOSITION OF GROUNDWATER IN
CRYSTALLINE ROCK AT GREAT DEPTHS
Analysis Unit Probable interval
Conductivity S/cm 400-600
pH 7.2-8.5
KMnC>4-consumption mg/1 5-10
Ca2+ mg/1 25-50
Na+ mg/1 10-100
Fe-tot mg/1 1-20
HCO3- mg/1 60-400
СГ mg/1 5-50
SO2" mg/1 1-15
F" mg/1 0.5-2
SiOj mg/1 5-30
HS' mg/1 < 0 .1-1
2.1. Thermodynamic aspects of copper corrosion
Copper is a relatively noble metal. Thus, the solubility of copper in pure
water is negligible. In the presence of certain reactants (oxidants) copper can
be oxidized and thereby attacked.
Thermodynamically, oxygen and sulphide have been found to be the
reactants which must be considered. Sulphate and nitrate are thermodynamically
possible oxidants but the reactions have a negligible rate in the absence of bacteria,
which are dependent on organic material for their life processes. Copper may
be attacked by oxygen in water solution if the redox potential is above -50 mV,
Standard Hydrogen Potential (SHE) at 25 —100°C (Fig.3). Thermodynamic
calculations have been performed for all the copper species that may be formed
in groundwater with an adequate chemical composition.
Oxygen may originate from air initially present in the pores of the buffer
bed. Oxygen present in the horizontal tunnel is supposed to be gettered by
iron(II) phosphate added to the sand-bentonite mixture. The very small amounts
of oxygen present in deep groundwater (< 0.1 mg/1) and formed by radiolysis
may be neglected.
According to the following equation:
2Cu + HS" + H20 ->■ Cu2S + H2 + OH"
508 EKBOM et al.
FIG.3. Equilibrium diagram for copper oxides and sulphides in water at 298K and 1 atm
pressure.
copper may be attacked by sulphide (HS~) in water, but only if the redox
potential is sufficiently low, i.e. below about —200 mV at 25°C or below —500 mV
at 100°C (Fig.3). The redox potential in groundwater is regulated by Fe(II) and
Fe(III) to a value around -200 mV.
The maximum rate of Cu2S formation is determined by the rate of supply
of HS-to the canister surface. Sulphide may be supplied from the groundwater.
Due to the presence of iron silicates in the rock the sulphide content is low
(0.1 mg/1). However, groundwater may contain up to 5 mg/1, probably as colloidal
metal sulphides which do not diffuse in the bentonite buffer. Conservatively,
however, calculations of the corrosion rate were made with the assumption that
this sulphide concentration occurred as SH“. Sulphide resulting from bacterial
reduction of sulphate may increase the sulphide content by not more than 2 mg/1.
2.2. Transportation of reactants to the canister surface
To determine the maximum corrosion rate of the canisters it has been
assumed that this is regulated by the transportation of reactants to the canister.
IAEA-SM-243/166 509
The formation of corrosion products, which may slow down the reaction rate
or the transport of corrosion products from the canister, was not considered.
It was furthermore assumed that all the reactants present contributed to corrosion.
Atmospheric oxygen, which is entrapped in the buffer, diffuses according
to calculations to the canister surface in less than one hundred years. Since
subsequently sources of oxygen may be neglected, sulphide will in the long term
determine the corrosion rate. Sulphide in the groundwater may be transported to
the canister surface in two ways (Fig.4), from cracks in the rock to the vertical
borehole or via the tunnel system down into the borehole.
The transport of sulphide from rock cracks through the buffer in the
borehole is assumed to take place only by diffusion (Dsooc = 4 X 10-10 m2/s)
as the permeability of the bentonite is extremely low (< 10-12 m/s). Diffusion
calculations were made according to a model developed by Neretnieks [5].
The transportation has three barriers to overcome:
the buffer in the borehole around the canister;
a zone of cracks in the rock around the borehole filled with swelling buffer
material; (this is the most effective barrier but has not been taken into
consideration in this calculation) ;
a zone of cracks in the rock around the borehole filled with water; (this
barrier has been found to be 10 times as great as the one first mentioned).
510 EKBOM et al.
According to the calculations the maximum amount of copper corroded
per year is 30 mg.
In calculating the sulphide transport from the tunnel down the borehole
the sulphide content at the mouth of the borehole towards the tunnel was
conservatively assumed to be the same as that in the groundwater in the rock.
Transport was calculated as a steady-state diffusion from the tunnel to the top
of the canister. This diffusion can give rise to a corrosion of the top of the
canister, which is less than 20 mg/year.
2.3. Type of corrosion
The corrosion attack will probably be of a general type with concentration
on the top of the canister, where the diffusion transport of sulphide to the
surface is greater.
The possibility of pitting, however, cannot be excluded. The relation
between pit depth (P) and time (t) may be expressed by the empirical equation
P = A(t-t0)n
where t0 is the incubation time until pitting starts, and A and n are constants.
According to corrosion tests in various types of soil in the USA over
14 years values of n were between 0 and 1, generally between 0.1 and 0.6 [6].
Further examination of these results has shown that n decreases with the time
of exposure. Thus the rate at which the pit deepens decreases with time and
eventually becomes negligible if the wall thickness is large enough. The continued
attack will then proceed mainly in the form of a widening of the pits or an
initiation of new pits. Thus, when the wall thickness is very large, as in this
case 200 mm, corrosion after very long times takes the form of an attacked
surface zone with local variations in depth.
The relation between the maximum depth and the average penetration is
called the pitting factor. In the field tests mentioned above, the greatest pitting
factor found for copper was 24. It decreases with the time of exposure. For
the canisters local variations have been taken into account by multiplying the
average penetration by a pitting factor of 25.
3. EVALUATION OF THE LEAD CORROSION
Reprocessed and vitrified waste from nuclear reactors will be disposed of
in titanium canisters with 6 mm thick walls and a 100 mm thick lead lining (Fig.l).
The canisters are disposed 500 m below ground in a way similar to that
described for the copper canisters. Two differences in the conditions are
IAEA-SM-243/166 511
essential to the corrosion rate. The canisters will be embedded not in compacted
bentonite but in a sand-bentonite mixture. No getter will be added to the
mixture of sand-bentonite in the tunnel, which means that entrapped oxygen
may contribute to corrosion.
3.1. Calculation of the rate of lead corrosion
If for some reason the titanium casing is penetrated locally, corrosion of
the lead lining can occur. It has been shown [4] that the corrosion rate in this
case, as in the case of copper canisters, is determined by the transport of
oxidants to the canister surface.
The maximum rate of corrosion can therefore be calculated in a way similar
to that for the copper canister. Oxygen in the vertical hole will be consumed in
a short time. Oxygen in the tunnel system, which in this case was not gettered
by iron(II) phosphate, diffuses with a maximum rate of 4 X 10-10 mol/s down
the vertical hole to the canister. This can cause a corrosion of not more than
5 g Pb/year. If, however, an addition of iron (II) phosphate had been made to
the tunnel buffer material this reaction would have been negligible and a much
lower corrosion rate, determined by sulphide diffusion, would have occurred.
3.2. Type of corrosion
It is supposed that the titanium casing is locally penetrated. The attack
on the lead lining thus has to start from a point and can spread out beneath the
titanium lining. It is supposed that the general shape of the attack will be semi-
spherical. The reason for this is that corrosion products will decrease the
diffusion and increase the electrical resistance within the attacked volume.
The penetration depth will thus depend on time, t as P ~ t1/3.
4. THE CORROSION RESISTANCE OF TITANIUM2
A comprehensive literature study and an inventory of experience compiled
in 1977 [7], confirmed that unalloyed titanium could be used in contact with
the expected environment of the depositories, without there being a risk of
local corrosion or hydrogen pick-up. The highest reported figure for general
corrosion was 6 /xm/year. This must be considered a conservative estimate,
because the corrosion rate used is at least a factor ten greater than other reported
values. The long-term exposure of titanium, which is reported here, was initiated
2 Corrosion tests on titanium have been reported on by S. Henrikson, M. de Pourbaix,
S. Poturaj.
512 EKBOM et al.
in the first place to obtain more certain general corrosion data, and in the second
place to verify its resistance to local corrosion and hydrogen pick-up.
In conjunction with study of the above-mentioned literature, the risk for
delayed failure in disposal canisters of titanium was considered [7, 8]. Whilst
the conclusion was that delayed failure could be prevented by maximizing the
initial hydrogen content of the titanium to 20 ppm, it was not considered
necessary to carry out experimental studies concerning its susceptibility to
that kind of failure. More recently, however, the possibility of this type of
cracking at even lower hydrogen contents has been discussed [9]. An experimental
study of this is under way [10].
4.1. Long-term exposure of titanium
At the beginning of the study it was assumed that the most corrosive
environment which could be present in the underground disposal sites would
arise from the leakage of Baltic sea water. It was therefore decided that the
exposure should be carried out in natural Baltic sea water (4000 ppm СГ) with
additions of NaF to a total F~ concentration of 10 ppm and acidified with HC1
to a pH of 4.5
Specimens for the determination of general corrosion by weighing, crevice
corrosion specimens, and specimens for use in demonstrating possible hydrogen
pick-up by slow tensile testing (1.4 X 10-4 min"1) and analysis for hydrogen,
were initially exposed for 300 days with the following temperatures and
oxygen contents:
(1) 100°C, < 10ppb02
(2) 100°C, 8 ppm 0 2
(3) 130°C, < 10ppb02
(4) 130°C, 8 ppm 0 2
In all four cases the specimens were exposed entirely in the liquid phase
and either half or entirely immersed in a buffer of 90% Si02 + 10% bentonite.
After the first 300 days the exposure in the first-mentioned environment
was continued for a further 300 days. This environment was considered to be
the most relevant when taking into account the most recent observations
concerning the chemistry of the groundwater. Since the actual temperature would
not exceed 70°C and the maximum СГ concentration is estimated at 400 ppm,
the experimental conditions must be considered to be much accelerated. In
reality the pH value would not sink below 7.2.
The test material was 1.5 mm plate of Grade 1 titanium, as received, and
with the following chemical composition:
IAEA-SM-243/166 513
0.003 0.006 0.07 14 0.04
The results of the exposure can be summarized as follows:
( 1) General corrosion or oxidation after 300 days was approximately 0.1 дт/уеаг,
and after 600 days it decreased to 0.1 дт/уеаг. This corresponds to a life
expectancy of tens to hundreds of thousands of years for canisters
6 mm thick.
(2) No local corrosion had occurred on the crevice corrosion specimens.
(3) The hydrogen content had not increased; neither could any significant
differences be observed in the results of the slow tensile testing.
4.2. Measurement of the corrosion potential of titanium during
gamma irradiation
The potential of Grade 1 titanium was measured against a saturated calomel
electrode both for 168 h during gamma irradiation at a dose rate of 0.4 rad/s,
and for 120 h after irradiation [7]. The experiments were carried out at room
temperature, and the titanium was immersed in the previously mentioned buffer
which had been doused with water of the following composition:
(1) Baltic sea water, pH 7.8, 8 ppm 0 2
(2) Baltic sea water, pH 7.8, 15 ppb 0 2
(3) Groundwater, pH 8.5, 8 ppm 0 2
(4 ) G roundwater, pH 8.5, 15 ppb 0 2
The potential of titanium was found to be completely unaffected by the
gamma irradiation. It stabilized at approximately +100 mV/SHE in environ
ment 1, +25 mV/SHE in environment 2, +100 mV/SHE in environment 3, and
+50 mV/SHE in environment 4. It can be seen from Fig.2 that these potentials
lie well within the passive region for titanium.
REFERENCES
[1] Final Storage of Used Nuclear Fuel, Kàrnbrànlesâkerhet, KBS Technical Report (1978).
[2] Reprocessed and Vitrified Waste, Kàrnbrànlesâkerhet, KBS Technical Report (1977).
[3] Swedish Corrosion Institute with reference group, “Copper as a canister material for
unprocessed nuclear fuel waste. — Evaluation from the corrosion point of view” ,
Karnbranlesakerhet, KBS Technical Report 90 (March 1978).
C% N% 0 2% H2 ppm Fe%
[4] Swedish Corrosion Institute with reference group, “Lead-lined titanium canister for
reprocessed and vitrified nuclear waste. — Evaluation from the corrosion point of view” ,
Kârnbrânlesâkerhet, KBS Technical Report 107 (June 1978).
[5] NERETNIEKS, I., Transport of Oxidants and Radionuclides through a Clay Barrier,
Report from Dep. Chem. Eng., Royal Inst. Techn., Stockholm (1978).
[6] DENISON, J.A., ROMANOFF, М., Soil corrosion studies, 1946 and 1948: Copper alloys,
lead and zinc, J. Res. NBS, Washington 44 (1950) 259.
[7] HENRIKSON, S., PETTERSSON, K., “The suitability of titanium as a corrosion resistant
canister for nuclear waste” , Kârnbrânlesâkerhet, KBS Technical Report 11 (1977) 35
(in Swedish).
[8] PETTERSSON, K., “Evaluation of the risk for delayed failure in titanium” , Kârnbrânle-
sákerhet, KBS Technical Report 26 (1978) 15 (in Swedish).
[9] PETTERSSON, K., “Complementary view-points on delayed failure in titanium” ,
Studsvik Technical Report MZ-78/116 (1967) 6 (in Swedish).
[10] PETTERSSON, K., Studsvik Energiteknik AB, private communication.
[11] HENRIKSON, S., DE POURBAIX, М., “Corrosion testing of unalloyed titanium in
simulated disposal environments for reprocessed nuclear waste. Final report” , Kërnbrânle-
sâkerhet, KBS Technical Report 96 (1978) 33 (in Swedish).
[12] HENRIKSON, S., POTURAJ, S., “Determination of the galvanic corrosion between
titanium and lead and measurement of the corrosion potential of titanium during
gamma irradiation” , Kârnbrânlesâkerhet, KBS Technical Report 67 (1978) 16 (in Swedish).
5 14 EKBOM et al.
DISCUSSION
K. KÜHN: After reading the KBS reports and having listened to your
presentation as well as that of Mr. Pusch (paper IAEA-SM-243/22), I now
wonder why you are still considering deep geological formations for disposal.
Why aren’t you in Sweden opting for shallow land burial or a man-made structure?
L.B. EKBOM: Perhaps Mr. Ahlstrom would like to reply?
P.-E. AHLSTROM: In order to meet the very stringent requirements of the
Swedish Stipulation Law, we think it prudent at this stage to have multiple
barriers. These barriers should have different functions and, to some extent,
be independent of each other in order ot provide some redundancy.
L.C.E. DEVELL: Supplementing Mr. Ahlstrom’s comment, I should like
to point out that the rock will also act as an effective barrier during glaciation
periods, in the course of each of which the earth surface may be eroded to a
substantial depth.
D.F. DIXON: Your test work is concerned with establishing useful life
with regard to corrosion. Experience with engineered components indicates
that components fail at a low random rate during the useful life. Have you
used the reliability theory to estimate failures during the useful life?
L.B. EKBOM: I shall ask Mr. Hannerz, one of my co-authors, to answer
that question.
IAEA-SM-243/166 515
A.К. HANNERZ: As regards the corrosion life of the canisters, the purpose
of the calculations in the paper is to establish the minimum life rather than
the “useful” life. This is done by:
(a) Assuming an ultra-pessimistically high concentration of corroding agent
(free sulphide ion). (In reality, free sulphide is unlikely to exist at the depths
under consideration because of conversion into iron sulphides due to contact
with ferrous silicates present on the surfaces of the crack system, but this was
disregarded. The absence of oxygen is proved by actual water sampling and
by the fact that ferric iron is consistently absent from the crack surfaces in
drilling cores).
(b) Calculating the maximum amount of corroding agent that could possibly
reach the canister surface, on the assumption that the above-mentioned high
concentration is constantly present in groundwater and on the basis of the
measured diffusivity values of the bentonite backfill (see paper IAEA-SM-243/108).
Special experiments have confirmed that the bentonite was not transported
away with groundwater. It will be converted from the sodium into the calcium
form on the time scale of millions of years, but this is not expected to increase
diffusivity.
(c) Assuming that the corroding agents will react with the canister surface as
soon as they reach it.
(d) Assuming the maximum pitting ratio (i.e. maximum to average corrosion
depth) ever observed in copper corrosion in soil (about 25). (Note that it may
be physically impossible for such a pitting ratio to occur here because thermo
dynamics requires the sulphide to diffuse to the actual corrosion site, unlike
the case of corrosion by oxygen. Transport limitation for the sulphide ion in a
deep pit would in fact soon stop its further penetration.)
By such a calculation we arrive at a minimum corrosion life of many
millions of years. The ultra-pessimistic assessment by the group of independent
scientists which was asked to consider the proposal includes a further safety
factor and reaches a final figure of “hundreds of thousands of years”.
In actual fact, the life-limiting phenomenon for this copper canister is
unlikely to be corrosion but rather eventual creep rupture due to the build-up
of internal helium pressure from alpha decay of the actinides in the unreprocessed
fuel. However, this will take millions of years.
In reality, therefore, we have a canister backfilling concept in which the
fuel will be isolated from groundwater for a time sufficient for the decay of
all significant radionuclides resulting from reactor operation (with the possible
exception of iodine-129). What is ultimately left to come into contact with
groundwater is uranium moved from the mine to the repository site.
S.R. HATCHER: In our recent drilling in granite at Whiteshell we discovered
groundwater with a chloride concentration of several thousand ppm at a depth
of a few hundred metres. What effect do you think such chloride concentrations
would have on the corrosion of lead and copper?
516 EfCBOM et al.
L.B. EKBOM: In the case of a copper canister, a high chloride content in
the water would increase the solubility of the copper. A chloride concentration
of several thousand ppm would mean an increase of two orders of magnitude in
comparison with the case of one hundred ppm. However, the solubility is still
so low ( 1СГ7 M) that the diffusion of the copper through the bentonite buffer
would be negligible. Probably the risk of pitting corrosion would increase slightly.
The situation with regard to lead would seem to be the same. The formation
of lead carbonates would effectively slow down the diffusion from a corroded area.
The risk of crevice corrosion on titanium might increase somewhat. The
conditions would still be far from those under which crevice corrosion has so
far been observed.
E.G. RAMO: What is the principal reason for selecting different canister
materials for vitrified waste and spent fuel?
L.B. EKBOM: The titanium-lead canister would appear to acquire an
appreciable corrosion lifetime, comparable to that of the copper canister, if
iron (II) phosphate is added to the buffer mixture of sand and bentonite as an
oxygen getter. This was not initially realized but can be concluded from our
corrosion evaluations.
P.-E. AHLSTROM: I should like to add that the amount of uranium and
plutonium in spent unreprocessed fuel is higher by a factor of about 200 than
in the vitrified high-level waste. This is one important reason why we considered
it prudent to use the more durable and more expensive copper canisters for the
spent fuel case.
P. FRITZ: In your analyses of canister stability you assume pH values of
7 to 9. At the Stripa test site measurements consistently give pH values above
9 and there are values as high as 9.8. In general terms, what effect would such
high pH values have on the stability of the type of canisters you investigated?
L.B. EKBOM: The high pH value will decrease the solubility of metal
ions and corrosion products. In the case of a copper canister, the reaction with
sulphite is improbable, and if no oxygen is present, the corrosion becomes negligible.
However, in our concept, with a buffer of bentonite the pH value will be
stabilized around 8.5.
CHAIRMEN OF SESSIONS
Session I A. BARBREAU
Session II I. S. ZHELUDEV
Session III N.S. SUNDER RAJAN
Session IV K. KÜHN
Session V P.A. WITHERSPOON
VOL.
Session VI H. STIGZELIUS
Session VII L.B. NILSSON
Session VIII V.I. SPITSYN
Session IX F.S. FEATES
Session X . E. MALÁSEK
Round Table C.A. HEATH
VOL. I
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