the pebble bed high temperature reactor as a source of
TRANSCRIPT
Jül -1114 - RGOktober 1974
KERNFORSCHUNGSANLAGE JÖLICHG E S E L L S C H A F T M I T B E S C H R A N K T E R H A F T U N G
Institut fur Reaktorentwicklung
The Pebble Bed High Temperature Reactor
as a Source of Nuclear Process Heat
Volume 2
Core Physics Studies
A Common Study by
Kernforschungsanlage Jülich GmbH
and General Electric Company
by
E.Teuchert, LBohl, H.J.Rütten and K. A. Haas
Als Manuskript gedruckt
ROERMOND:
KREFE
ELSDORIwÄRERGHEIM
00
Berichte der Kernforschungsanlage JDIidi - Nr. 1114
Institut für Reaktorentwicklung Jöl - 1114 - RG
Dok.: Pebble Bed High Temperature Reactor - Nuclear Process HeatNuclear Process Heat - Pebble Bed High Temperature ReactorCore Physics Studies
Im Tausch zu beziehen durch: ZENTRALBIBLIOTHEK der Kernforschungsanlage Jülich GmbH,Jülich, Bundesrepublik Deutschland
The Pebble Bed High Temperature Reactor
as a Source of Nuclear Process Heat
Volume 2
Core Physics Studies
A Common Study by
Kernforschungsanlage JOlich GmbH
and General Electric Company
by
E. Teuchert, L Bohl, H. J. Rütten and K. A. Haas
ACKNOWLEDGEMENT
The authors wish to express their gratitude to Mr. Lou Woikefor his support in the economy study, and to Mrs. A. Conrad-Wienands for her cooperative aid in preparing the manuscript
Kernforschungs-anlage Jülich JÜL - 1114 - RG Oktober 1974GmbH IRE
THE PEBBLE BED HIGH TEMPERATURE REACTOR
AS A SOURCE OF NUCLEAR PROCESS HEAT
Volume 2
CORE PHYSICS STUDIES
ABSTRACT
A theoretical analysis is given for a series of 8 different vari-ants of the pebble-bed reactor in the "once through" fuel manage-ment scheme. The comparison brings the insight into the parametricsensitivities and into the development potential of this type. Thethorium/U-233 recycling fuel cycle allows to increase the conver-sion ratio up to the range between 0.90 and 0.95. The feasibilityis outlined for changing over between different fuel cycles underfull power operation. - The study is complemented by a review ofthe relevant previous researches.
CONTENT
Page
ABSTRACT O
SUMMARY AND CONCLUSION
O.I0.20.30.4
IntroductionSummaryConclusionsRecommendation
1• 1
25
1 REACTOR DESIGN CONCEPT AND ANALYSIS 7
1.1 Introduction 71.2 Core Physics 71.3 Fuel Cycles 91.4 Fuel Element Design • ~ 101.5 Core Analysis Model 101.5.1 Design and Life Simulation . 101.5.2 Calculational Methods 111.6 Design Constraints 13
2 SELECTED DESIGN STUDIES 15
2.1 Introduction 152.2 Parametric Specification 152.3 Characteristics of the Equilibrium Cycles 192.3.1 Variation of Burn-up 212.3.2 Variation of the Gas Outlet Temperature 322.3.3 The High Performance Variant 332.3.4 Low Power Density 342.3.5 Low Enriched Uranium Versus High Enriched Uranium/Thorium
Fuel Cycle 342.3.6 Low Power Density 362.3.7 High Conversion Rate 372.4 Sensitivity Studies 392.4.1 Heat Transfer Variations 392.4.2 Cost Parameter Variations 422.5 Consideration of Hot Spot Analysis 42
3 CHANGING-OVER IN FUEL CYCLE 44
3.1 Introduction . 443.2 Different Fuel Cycles 443.3 Switching-over between them 473.4 Mass Balances 483.5 Fuel Cycle Economy 49
ii
Page
4 PREVIOUS RESEARCH ON TRENDS 52
4.1 Introduction 52
4.2 isotopic Distributions in the Core 524.3 First Core and Running-in 544.4 Influence of the Core Height 554.5 Variational Studies . 564.6 Manipulation of the Radial Power Density Profile 584.7 Fuel Cycle Comparison 594.8 Ball Flow Distribution 594.9 Temperature Distribution in the Core 6O4.10 Thermal Conductivity 624.11 Fast Neutron Dosis at Reflectors 644.12 Effect of the Upper Void . 654.13 Reactor Control . 664.14 Reactor Shut Down 674.15 Xenon Distribution 694.16 Temperature Coefficients 70
REFERENCES 72
i i i
LIST OF FIGURES
Fig. 1 Distribution of the fissile isotopes in the core ' 8"Fig. 2 Neutron flux distribution for 2-zone core 8Fig. 3 Power distribution 8Fig. 4 Axial distribution of power density and temperatures in
a shell ball variant 9Fig. 5 Temperature distribution in conventional and shell ball design 10Fig. 6 Model of the numerical reactor simulation ' 11Fig. 7 V.S.O.P. Code System 12Fig. 8 Conversion and fuel cycle costs as function of burn-up 31Fig. 9 Temperature profile under varying power density 41Fig.10 Switch-over in fuel cycles 47Fig.11 Mass and cost balance for sequence of fuel cycles 48Fig.12 Axial distribution of macroscopic cross sections 53Fig.13 Local distribution of silver isotopes 53Fig.14 First core loading 55Fig.15 Running-in period 55Fig.16 Axial power density depending on core height 56Fig.17 Power density profile depending on fuel cycle and moderation ratio57Fig.18 Power density profile for various designs 58Fig.19 Power density for varying 2-zone loading 58Fig.20 Varying ball flow pattern and corresponding burnup dis-
tribution 60Fig.21 Power and temperature distribution for variation of gas out-
let temperature 61Fig.22 Thermal loads for conventional and shell ball 62Fig.23 Fraction'of fuel above given temperature 62Fig.24 Fraction of fuel at given temperature and burn-up 63Fig.25 Thermal Conductivity in axial positions in the core 63Fig.26 Fast neutron flux at the edges of the core 64Fig.27 Fast dosis vs. temperature at side reflector in 30 years 64Fig.28 Simulation of the area above the core for transport •
calculation ' 65Fig.29 Thermal flux variation when inserting control poison into
the upper cavity 65Fig.30 Change in power and temperature distribution at withdrawal of
control poison 66Fig.31 Thermal flux in r-© coordinates
A) Unrodded core; B) rods inserted 68Fig. 32 Efficiency of shut down rods bank 68Fig.33 Xenon Override 69
i v
LIST OF TABLES Page
Global DataMass BalanceFuel.Cycle CostsHeat Transfer
I Reactor Parameters,II Fuel Cycles at Equilibrium •. .III Range of Technological ExperienceIV Descriptive Names of Explicitely Calculated CasesV Characteristics of the CasesVI Core Design Parameters .VII Fuel Element ParametersVIII Cost ParametersIX Results for Equilibrium CoresX Results for Equilibrium CoresXI Results for Equilibrium CoresXII Results for Equilibrium CoresXIII*a Key to Tables XIII, b-iXIII,b Case 1013 ' -sXIII,c Case 1113XIII,d Case 1213XIII,e Case 1313 .XIII,f Case 1433XIII,g Case 4011XIII,h Case 4021XIII,i Case 9022 . .' .' .XIV Neutronic BalanceXV Power Density Variation at Case 4021
(Gas Temperature Rise 250 - • 800 °C)XVI Variation of Reactor Power OutputXVII Sensitivity of fuel cycle costs on cost assumptionsXVIII Design data (1000 MWth) . .XIX Cost data ..assumptionsXX The four different equilibrium cyclesXXI Break down of fuel cycle costs (mills/kWhe) ,XXII Fuel cycle costs (mills/kWhe) under cost data variationsXXIII Fuel element identification for the first coreXXIV Shut-down reactivity changes
141316161717181920202122232425262728293038
40414244454649505567
0 SUMMARY AND CONCLUSION
0.1 Introduction
The objective of this volume is to examine the detailed nuclear/thermal/fuel-
cycle cost performance of a selected number of OTTO-cycle pébble-bed reactors
for process heat/electricity application. Other objectives are to assess the
sensitivity of the reactor performance to variations in the design, operation-
al, and cost assumptions, and to discuss considerations affecting the design
and operation of the reactors. The result is a survey of the characteristic
properties of this type of reactor, in which the mutual coupling between the
continuously moving fuel and the local distribution of burnup and power pro-
duction is the most typical feature.
0.2 Summary
A parametric study was made, varying the reactor parameters shown in Table I.
The conclusions with respect to each of these variations is presented below.
TABLE I: .Reactor Parameters
Reactor power
He inlet pressure
He inlet temperature
He outlet temperature
Power density
Core height
Burnup
Fuel cycle types
Fuel element types
Moderation ratio
Coated particlediameter
MWth
at
°C°CMWth/m3
m
MWd/t
VNHM
um
3000 - 6000
40
25O
800 - 1190
5 - 12
3.75 - 6.00
36000 - 130000
LEU: Low enriched uranium
Th/U5: Th/high enriched uranium feed
Th/U3: Th/U-233 feed, recycle
Conventional ball
Zoned shell ball
110 - 380
400 - 800
2
(section 0.3). Selected design cases formed by appropriate combinations of
the reactor parameters were studied with respect to the neutronics, thermal
and cost performance. The results, presented in section 2, illustrate the
design flexibility of the OTTO cycle concept over a wide feasibility range.
Of particular importance, is the possibility of utilizing either the low-en-
riched uranium or the thorium fuel cycle, including the ability to have a
smooth on-line transition between these cycles. Moreover, with the recycling
öf the U-233 from the thorium cycle, it is shown that a very high conversion
ratio (~ 0.95) can be obtained involving a certain increase in the fuel cycle
costs. '• '
The OTTO cycle concept and the methods of analysis used for this report are
described in section 1. A discussion of the on-line transition from one fuel
cycle to another" is given in section 3. In section 4, previous work on the
OTTO cycle is surveyed to provide a basis for understanding the trends used
in arriving at balanced designs. Extensive references are provided.
0.3 Conclusions
The following statements result from the representations of section 2 through
4 and from previous parametric studies:
•* Regarding the fuel element-design constraints, the OTTO concept allows
to provide an average power density of 12 MW/m , average gas outlet tem-
perature of 119O °C, and average burnup of 130,000 MWd/t, simultaneously
* Reduction in the power density from 12 to 9 or even to 5 MW/m at a
simultaneous increase of the core volume «is advisable.
- It brings a high flexibility for optimizations with respect to the de-sign of core and fuel elements, fuel cycles, thermal loads, control,safety, fuel cycle costs, and requirements of application.
- It increases the fuel cycle costs only by less than 5 %..- It improves slightly the .neutron economy.- It increases the conversion ratio (uranium cycle*%* 5 %, thorium cycle*s,ll % ) . . .
- It lowers significantly the core pressure drop.- One and the same core design can be used for plants with different to-
tal reactor power output (e.g. 3000 MWth at 5 MW/m3 — ^ ÔOOOMWth at10 MW/m3).
The fuel element design should depend on the requirements of fuel cycle
and thermal hydraulics.
- High resonance escape probability is achieved by the choice of big coat-ed particles, lumped matrix, low heavy metal content per ball, admixingof pure graphite balls in the core charges.
- Separation of feed and breed fuel in different ball types brings a flexi-bility for fabrication and reprocessing.
- Radial zoning of the matrix (shell ball design) decreasesconsiderablythe peak temperature at the center.
# The choice of the core dimension requires the trade-off between many para-
meters.
Increasing the core height from 5 to 6 m:
- increases the peak/average ratio of the axial power distribution almostproportionally.
- Increases the core pressure drop proportionally.- Increases the required insertion depth for the shut-down rods.- Reduces the perturbation on upper-core ball flow distribution arisingfrom the ball flow towards the disloading tubes at the bottom.
- Affects only slightly the neutron economy.
Increasing the core radius under constant height:
- Leaves the axial power profile almost constant.- Leaves the neutron economy almost unchanged.- Influences significantly the ball flow trajectories.- Requires possibly a modification in the design of fuel element dis load-ing devices (1 - 3 - 7 disloading tubes).
* An increase of the reactor coolant temperature rise to 1190 °C is poss-
ible from the view of thermal and neutronic performance.
- It increases the fuel element temperature a like amount.- Decreases the required core pressure drop significantly.- Only weakly affects the core nuclear performance.- Effect on fuel cycle costs is in the range of 1 - 2 %.- It does not require any reduction in the local power peaks.- It does not require any reduction in burnup.
* Reactor control can be performed by moving the control rods in the cavity
between the pebble-bed and the top reflector.
- During control adjustments there is no significant excursion of localpower or temperature in the fuel elements.
- Reduction of the fast neutron dosis at the top reflector graphite isautomatically achieved by the presence of the control system.
- Cooling of the control rods is automatically provided by the inlethelium.
Different fuel cycles are feasible and economical (Table II).
- The thorium cycle has higher conversion ratios, particularly withÜ-233 recycle.
- The thorium cycle has a 10 - 20 % flatter axial power distribution thanthe uranium cycle.
TABLE. II:Fuel Cycles at Equilibrium
Ball type
Moderation ratio
Coated particlediameter
Conversion rate .
Power peak/avg.
Fuel cycle costs
VNHM
pn
mills/kwh
Low enricheduranium
shell
360
800
0.52-0.58
2.6 - 3.2
2.2 - 2.6+)
Th/U-235feed
conv.
240
600
" 0.58 - 0.
2.2 - 2.
~ 2.1
63
6
Th/U-233feed, recycle
conv.
110
400
0.94
1.7
2.9+)
Strongly reduced burnup.
- The fuel cycle costs are slightly lower for the thorium cycle, exceptfor the recycle design which was selected for high conversion ratiosat the expense of fuel cycle costs.
- Central fuel element temperatures in the uranium cycle case are com-parable to the thorium case because of the shell ball.
The low enriched uranium fuel cycle requires to select the fuel element
design in view of high resonance escape probability (N_,/N > 3OO) .
- The fuel element burnup affects the fuel cycle costs and conversion ra-tio as shown in Fig. 8.
- Fuel cycle costs approach a minimum at burnups being higher than con-sidered in the design cases.
- The reduced heavy metal content involves a relatively high contributionof fabrication and reprocessing in the fuel cycle costs.
- There is an upper limit for the conversion ratio in the range between0.60 and 0.65.
- The incore utilization of the bred fissile plutonium is in the rangebetween 85 and 90 %.
- The revenue for disloaded plutonium is low due to the presently lowequivalence value.
The highly enriched uranium/thorium fuel cycle is marked by a very high
flexibility for variations in fuel element design and fuel cycle adap-
tion.
- The carbon/heavy metal ratio should be Nc/Nmi > 1°° f o r keeping the
neutron energy spectrum soft.- Recycling of the bred U-233 improves the economy of neutrons and fissile
material.
- Recycling basing on a feed-breed concept allows to separate undesiredisotopes (e.g. U-236, Pu) from the fuel cycle.
- Feed and breed fuel can be inserted in different balls which can bemechanically separated before reprocessing.
- Improvement of conversion rate can be achieved by low Nc/Ngjj ratio, bythe use of pure U-233 as feed fuel, low burnup, large core configura-tion.
- On principle the conversion rate 1.0 is achievable.- At high conversion (CR = 0.94) the fuel cycle costs are by 1.5 higher
than at optimum due to the increase in fabrication, reprocessing, andworking capital costs.
- Increase of the conversion flattens the axial power profile.
One and the same reactor design can be operated in various fuel cycles.
- The efficiency of the control system fulfills the varying requirements.- Change-over between the cycles can be performed under full power opera-
tion.- Power density profile changes smoothly into that of thé subsequent fuel
cycle.- Change-over from uranium to thorium cycle requires to submit an inter-mediate buffer layer of balls with reduced fissile concentration.
- From time to time the utility, which operates the reactor, can adaptthe fuel cycle to changing requirements of the market.
O.4 Recommendation
* The next steps of the reactor physics research should be the assessment
of one or few conceptional guide designs and an in-depth study.
* The assessment of the core configuration requires the trade-off between
the results of this study and the relevant engineering aspects. These
are:
- Design optimization of the core bottom, i.e. number and location ofdisloading tubes.
- Determination of the corresponding ball flow trajectories.- Insertion depth of the shut-down rods.- Balancing power density vs. emergency heat removal.- Decision between fixed or replacable reflector.
* The assessment of the fuel cycle requires the trade-off between the re-
sults of this study, the aspects of the fuel element technology, the
expected development of the reprocessing techniques, the conservation
of the fuel resources, and the economy. Desirable work is:
- Research of various recycling strategies for the thorium cycle by meansof computer simulation, and optimization.
- Optimization of realistic recycling variants for conversion rates in
the range between 0.85 and 1.00.- Development and testing of fuel elements with high heavy metal content.- Fuel element testing at the thermal load condition of process heat react-or application.
- Cheapening of the fabrication.
Each conceptional design requires the in-depth research, which is expect-
ed to bring further aspects for the choice of a final reactor concept.
The items are:
- The many currently known ones (running-in, control and shut-down per-formance, xenon stability, hot spot analysis, temperature coefficients,accidental analysis, fission product release, etc.).
- Three dimensional simulation of the control rods in the cavity abovethe balls.
- the effect of the uncertainties, being inherent to the statistics ofthe pebble-bed.
- Completion and testing of the computational tools.
The final concept should make use of the possibility for flexible changes
in the fuel cycle being guided by the completion of the outer part of the
cycle, by resource conservation, and by economy.
REACTOR DESIGN CONCEPT AND ANALYSIS
1.1 Introduction
A recently developed variant of the pebble-bed HTR is the OTTO operating
scheme /I/. The fuel balls are not circulated several times through the
core as done in the AVR and THTR, but they pass very slowly only once from
the top to the bottom. The helium coolant streams parallel to the balls,
and the age dependent power production is by this way well balanced with
the rising temperature in the helium. This scheme minimizes the peak fuel
temperature and, therefore, it allows an increase of the gas outlet tem-
perature up to 1150 °C without any necessity for reduction of the average
power density and of the fuel burn-up, which has already been tested up to
16 % FIMA in the AVR.
In this variant of the pebble-bed reactor there is a strong mutual coupling
between the local distribution of neutron flux, fuel depletion, and pattern
of the ball flow velocity. Variations in the design parameters of the core
and of the fuel elements bring not only the currently discussed changes in
fuel cycle costs, but they bring also marked variations in the power dis-
tribution in the core and by this way in the thermal loads on the fuel ele-
ments. These variations are more pregnant than in reactors with mixedly in-
serted fuel elements, therefore, optimizations must be based on the know-
ledge of the various tendencies. Before going into the detailed research
of these trends this chapter will briefly outline the basic feature of this
reactor.
1.2 Core Physics
Due to the continuous movement of the balls there is a marked gradient of
the fissile content in the balls from top to bottom. Fig. 1 gives the dis-
tribution of the depleting U-235 and of the emerging U-233 for a near to
optimum design in the thorium fuel cycle. A two zone fuelling device pro-
vides a higher enrichment for the balls being inserted close to the outer
radial core area, which achieves an equalization of the axially integrated
power and, simultanously, a flattening of the radial gas outlet temperature
profile. The use of a lower moderation ratio N_/N in the outer zone allowsC HM
to flatten the power peak in the upper core area (/3/ Chapter 3.7).
FISSILE ISOTOPES
U-235
be seen the well-known superelevation
close to the core region. In the outer
radial core zone the thermal flux is
slightly reduced, which is due to the
higher absorption by the higher heavy
metal loading of the balls. The shape
of the fast flux is very similar to that
of the power density which is given in
Fig. 3.
POWER PER BflLL
FIG. 1:Distribution of the fissile iso-topes in the core.
Fig. 2 presents the corresponding
distribution of the neutron flux.
The thermal flux is formed by the
balance between the local neutron
supply rate from the epithermal
energy range and the thermal ab-
sorptions. For the reflector areas
the flux is given by the thin
lines in the drawing, and it can
FAST FLUX
BOTTOM
FIG. 2:Neutron flux distribution for2-zone core.
FIG. 3:Power distribution
OM
POWER HNO TEMPERATURES PT THE RXIS OF THE CORE(LOW ENRICHED. 8MW/M«*3. 115000 HWD/TI
CELSIUS1S00
S.7 =TEST EXPERIENCE POWER
»ERIENCE FUELTEMP.
100 200 300TOP
400 500 cnBOTTOM
FIG. 4: Axial distribution of power density and temperaturesin a shell ball variant.
Fig. 4 gives an outline of the coupling of the axial power distribution and
the heat transfer. It shows that variations in the maximum power - within
a certain range - do not affect the peak fuel temperature. Herefrom results
that an increase in the envisaged burnup, which shifts the asymmetric shape
of the power more towards the top, is not limited by a restriction from the
viewpoint of heat transfer, which is a rather unusual phenomenon in reactor
core physics.
1.3 Fuel Cycles
This type of reactor can be operated in the low enrichment fuel cycle /2/,
in the highly enriched uranium/thorium fuel cycle /3/, and in a high con-
version thorium cycle as well /16/.. Furtheron it can easily be switched over
from one cycle to another under full power condition so that the operating ,.
utility has the freedom of adapting to the most economic fuel cycle at any
time /I7/. As soon as reprocessing and refabrication will be developed for
the large scale application, the high conversion cycle is supposed to be-
come very attractive, for it helps to improve the economy of the primary
energy sources. The potential for producing high gas temperatures allows
to convert the nuclear energy into energy forms other than electricity at
a considerably higher thermal efficiency.
1.4 Fuel Element Design
10
COMVENIJONflL BfiLL
SHELL BfiLL
•
800-
600-
•
400-
200-
TEMPERflTURE PROFILECELSIUS
CONVENT.BflLL
iSHELL BflL
L NS
_ËfiS_
2 3 CM
PäHERi 2 .4 KH/Bfl'.LDOSISi 1 .8«10 2 1 N/CH2
FIG. 5; Temperature distribution in conventionaland shell ball design.
The fuel elements are graphite balls of 6 cm diameter which contain a sta-
tistical distribution of coated particles in an inner spherical region. In
a new variant, the coated particles are located in a spherical shell of 1 cm
thickness surrounding a globular heart of pure graphite with a diameter of
3 cm (Fig. 5). By this shell ball design the temperature load is strongly
reduced due to the shorter distance between the coated particles and the
surface of the ball. The total temperature difference in the fuel region
is reduced by 56 %. This fuel element allows a considerable increase of the
core power density without affecting the potential for achieving very high
gas outlet temperatures.
1.5 Core Analysis Model
1.5.1 Design and Life Simulation
The left hand side of Fig. 6 is a schematic draft of the reactor core. Here,
the dotted curves represent the stream lines of the balls as given by ex-
11
MODEL OF THE PEBBLE BEO HEHCTOR CniCULflTIONBL ngQEL
M M I t
TOP REFLECTOR RMO VOID
BOTTON REFLECTOR
FIG. 6:Model of the numerical reactorsimulation.
periments. In the area close to
the radial reflector their velo-
city is considerably lower than
at the center line of the core.
The right hand figure gives the
calculational model. This ne-
glects the conical shape of
the core at the bottom, and at
the top, the void and the per-
forated reflector are replaced
by a diluted graphite reflector
area. The degree of dilution has been adjusted by means of a comparison to
a 2-dimensional transport calculation, which allows a rigorous treatment of
the cavity above the bed of balls /8/. The core is subdivided into several
cylindrical zones, in which the balls move stepwise down. The different
speeds of the balls are simulated by a finer division of the steps in the
outer cylinders.
When the fuel balls are travelling slowly through the core, the depletion
proceeds due to the local neutron flux distribution. This distribution is
strongly coupled to the k of the balls which decreases corresponding to
the burn-up. To account for this mutual coupling, the computational proce-
dure follows explicitely the flow scheme of the fuel elements through the
core. This is done by following the reactor life from the first core load-
ing through the running-in period to the equilibrium fuel cycle under fre-
quent iterations on neutron flux and spectrum calculation.
1.5.2 Calculational Methods
The computations are performed by means of a recently developed code system
(Fig. 7), which allows the reactor life history from an initial loading:
through the running-in period to the equilibrium cycle to be followed /20/.
12
v, s. o. p.(SCHEMATIC)
I FUEL ELEMENTS] |CORE DESIGN
CORE NEUTRONICS. CONTROLBURN-UP
IN-DEPTH RESERCH:2D - TRANSPORT-CALCULATION.2D - CONTROL OF XE-TRAHSIEMTS
JD - CONTROL ROD INSERTIONFUEL ELEMENT LOAD HISTORYFISSION PRODUCT RELEASEI'.EFLECROR LOADS AND DESIGN
FIG. 7; V.S.O.P. Code System
The neutron spectrum is calculated by a combination of the GAM and THERMOS
codes. They can simultaneously be applied for many core regions differing
in temperature, burn-up, and fuel element lay-out, if required. The thermal
cell code THERMOS has been extended for the grain structure of the coated
particles inside the spherical fuel elements /15/, and the epithermal GAM
code uses modified cross-sections for the resonance absorbers prepared from
double heterogeneous ZUT-DGL calculations /I4/.
A fast one-dimensional diffusion code in four energy groups synthesized a
two-dimensional neutron flux map by means of r-z iterations. This is used
for the bum-up calculations in all core regions drawn in Fig. 6. The build-
up history of 43 fission product nuclides in these regions is followed ex-
plicitely. The diffusion part of the program system will be repeated at many
short burn-up stages, and the spectrum module will be re-iterated at some
larger time steps, when some significant change in the spectrum is expected.
The fuel management and cost module performs the fuel shuffling and general
13
evaluations, and, furthermore, processes the calculational results for follow-
ing more detailed physical investigations.
Apart from a number of newly developed partial programs, the code system is
based on a number of currently used computer codes as the ZUT, TUZ, GAM,
THERMOS, GATHER, FEVER, EXTERMINATOR are. Therefore, we call it the V.S.O.P.
system, which means "Very Superior Old Programs" /20/. It is able to follow
a reactor life history of 5 years with 30 spectrum runs and 150 diffusion
calculations in approximately 20 minutes of CPU time on the IBM 370/168.
1.6 Design Constraints
The fuel element components have been tested for a certain interval of de-
sign data. The range being used in current researches is given in Ref. /3/
(Chapter 2.2) and /4/ (page 4). In Table III the comparison of presently
established data and those being under research outlines the short term de-
velopmental potential of this reactor. The third column gives the corres-
ponding data for the reference case no. 1013 of the study in the subsequent
chapter. The distance from the data of the first column shows that the re-
ference case represents very conservative design. It leaves a wide margin
TABLE III:Range of Technological Experience
Radius of fuel kernel
Volume fraction ofcoated particle inmatrix
Maximum burn-up
Maximum fast neutrondosis
Maximum pwer per ball
Maximum temperaturein fuel
cm
MWd/t
1021/cm2
kW/ball
°C
Tested
0.03
0.20
120 000
8
5.7
1250
Development
0.04
0.30
140 000
9
6.8
1350
Ref.Case 1013of this study
0.04
0.09
116 000
4.6
4.9
1O74
14
for engineering factors and for further optimizations. In order to demons-
trate the safety margin for the fuel temperature, it has to be mentioned
that the coated particles are exposed to 1900 °C during their fabrication,
while the nominal value for the maximum temperature is only 1074 °C in the . •
reference case. Fig. 27 (below at page 64) gives the fast neutron do-
sis constraints for the reflector graphite. Depending on power density and
on core design, the curve of "swelling to the original dimension" is reached
after a time between 12 and 24 full power years.
15
SELECTED DESIGN STUDIES
2.1 Introduction
The intent of this chapter is to explore the feasibility range of the OTTO
concept by studying the reactor performance with several choices for the
major design parameters. The results presented include both neutronic and
thermal performance data and the corresponding fuel cycle costs.
2.2 Parametric Specification
The basic reactor data for the parameter study are shown in thé Tables I and
IV through VIII. The major design parameters varied in the study were power
density, outlet temperature, fuel cycle, fuel element and burn-up, as shown
in Table I. The Tables IV and V identify the selected bases for the para-
metric study, and the Tables VI through VIII give explicitely the data of
core, fuel element, and cost assumptions. In addition, for some of the cases,
the cost assumptions were varied to establish sensitivities (see 2.4.2).
Also, for one case, the reactor power was varied to determine the effect
on fuel element temperatures (see 2.4.1).
The series of explicitely calculated cases (Tables IV, V) has been chosen
to answer the major questions on the OTTO concept. The choice of the first
case 1013 bases on the knowledge of previous studies. The performance data
turn out to fit well into the range of present technological test experience
(Table III), and the fuel cycle costs are close to an optimum of the low en-
richment fuel cycle. The choice of the low enrichment cycle is a reasonable
solution as long as the closure of the thorium cycle is still under develop-
ment.
The variants 1113 and 1213 outline the influence of burn-up. The aim of case
1313 is to show the influence on the reactor performance given by an increase
of the gas outlet temperature. Case 1433 gives a core in which the maxima
of the fuel element power and temperature reach the limit of present test
experience. For comparison the case 4011 gives the highly enriched uranium/
thorium fuel cycle. Here, case 4021 shows the same fuel elements in a larger
16
TABLE IV:Descript ive Names of Exp l i c i t e ly Calculated Cases
I d e n t i f i c a t i o n
Base case, low enriched uranium cycle LEU
Low burnup variant
High burnup variant
High outlet temperature variant '
High performance variant
Variant in high enriched U-Th cycle Th/U5
Low power density Th/U5 variant
High conversion Th/U3 fuel cycle
PR-3000, previous LEU design
Label
1013
1113
1213
1313
1433
4011
4021
9022
23502
TABLE V:C h a r a c t e r i s t i c s of the Cases
Label:
Reactor power HWthAverage power density MWth/m3
Core height m
Fuel cycle type
Fuel element type
Core outlet temperature °C
Approx. burnup MWd/kg
1013 1113 1213 1313 1433 4011 4021 9022 23502
3000
9
5.0
12
3.75
LEU
Shell
9S5
100 70 130
1190
100 130
9
5 .0
5
6 .0
Th/US Th/U3
Conv.
985
100 40
5 .0
LEU
Shell
978
100
The nominal reactor mixed outlet temperatures are 960 oc and 1160 °C, respectively
core with the power density reduced to 5 MW/m , which is a very attractive
variant from the view of safety and component design. The last case 23502
is a previously derived LEU design of low power density /5/.
It should be noted in Table II that the thorium cycle reactors use the con-
ventional ball fuel element, while the shell ball was chosen for the low-
enriched uranium cycle reactors because its better temperature profile helps,
to offset the effect of higher axial power peaking. Further, the shell ball
design helps to improve the resonance escape probability by its higher lump-
ing of the fuel, which is a fundamental demand for the low enriched cycle
in contrast to the thorium cycle.
17
In detail the design data of the different cores are presented in Table VI.
For an OTTO pebble-bed there are several restrictions for the height of the
core which are different from those of other reactors. For the low enriched
cycle the height should not exceed 5 m from the view of the axial power peak-
ing. For the thorium cycle this restriction is about 5.5 - 6.0 m. Further
restrictions are given by the coolant pressure drop being dependent on the
average power density, and by the feasibility of pressing shut-down rods
TABLE VI iCore Design Parameter»
Cases
CorePower density
Height
Radius, inner/outer fuelling zone
No. of equlvolume channelsin inner/outer cone
Relative residence time in thechannelsinner zone
outer zoneBe inlet pressure
ReflectorThickness of top/bottom/radial R,Graphite density
top/bottom/radial R.
Homogeneous control poisonin top R. £ a
MH/m3
D
m
at
m
gr/cm
cm
1O13, 1113 , 1213 ,1313 , 4011
9
5 . 0
3.76/4.61
4 / 2
12/12/12/12
14/18
2 . 0
0.32
1433
12
3.75
4021, 9022
5
6 .0
4.98/5.64
23502
5
5 . 0
5.45/6.18
4 / 2
12/12/12/12/12/12/12
14/1640 .
/ 1.5 / 1.0
/ 1.60 / 1.6O
3.14-10"*
TABLE VII:Fuel Element Parameters
C a s e s
Coated particle fuel:
Kernel diameterThickness inner/outer coatingKernel densityDensity inner/outer coating
Fuel elements:Ball diameterInner/outer radius of matrixGraphite densityGraph, thermal conductivityBall packingReactor radial zones:Moderation ratioVol. fract. coat. part, in matrixHeavy metal loadingHH density in matrix
urnJim
g/cm3
g/cm
Cflt
cm
o/cir
W/°C-cmbal l s /m 3
"C^HM
g/ba l l
g/cn.3
1O13, 1113, 1213,1313, 1433, 23502
Low-enriched UOj
80O
110/809.501.00/1.85
Shell ball6
1 .5 /2 .5
1.70
Function (temoerature5394
inner3 8 0
0.074
9 . 9
0.19
outer3 2 0 •
O.Ö8711.70.23
4011, 4021
Th/O-235mixed oxide600
80/809.501.00/1.85
conventional60.0/2.51.70
, dosis) ranging5394inner250
0.09«
14.S
0.22
outer2 3 0
0.10415.8O.24
9022
Th/U-233
mixed oxide4 0 0
50/8O9.501.00/1.85
conventional6
0.0/2.51.700.17 - 0.395394inner outer
110
0.26732.40.50
18
into the bed of the balls. On the other hand a relatively great height helps
to equalize the flow pattern of the balls and to minimize the neutron leakage
losses. The conceptual data in Table VI represent a trade-off between the
various influences. .
The subdivision of the core into burn-up and fuelling zones (cp. Fig. 6)
is based on flow pattern measurements at small models, and further on the
accuracy demand for the expected results. In order to simulate the control
rods in the upper void a certain amount of thermal poison has been admixed
to the top reflector, which ensures the current reactor control including
the 1ÖO — > 40 % xenon override.
In Table VII it can be seen that for each group of reactors there are two
types of fuel elements, corresponding to an inner and an outer radial fuell-
ing zone. The balls of the outer zone have higher heavy metal loadings,
i.e. higher K . The purpose of this zoning is to flatten the radial depen-
dence of the outlet gas temperature as will be described in Chapter 4.6.
TABLE VIIIICost Parameters
U3O3 ore
U,0 ——* UFg conversion
Enrichment costs
Tail enrichment
E.g. U-235 at 93 »
U-233
PU-239.+ Pu-241 .
Th-232
Fabrication of coated particles
Manufacturing of. balls
Head-end, shipping
Reprocessing
Storage of waste (e.g. thorium)
Loss in fabrication and reprocessing
Annual load factor
Net electrical efficiency
Interest rates
Tax rate on fissile material
Lead time« payment of ore
Lead time, payment of enrichment service
Lag time, credit for discharged fuel
lag time, payment of reprocessing
*/ib u 3o 8
*/kg u
2/suti
*Ag"
n-
*Ag HM$/ball
*Ag c*/kg HM
*/kg HM
%
%
»
d
d
d
d
10
3.4
40
0.4
14 820
18 000
6 OOO
10
220
3.44
10.4
92
0.0
0.0
0.8
0.4
10
2
219
219
360 .
309
For the high conversion
variant 9022 this zoning
is-performed by grading
the enrichment only, because
a maximization of the heavy
metal in the core improves
the conversion rate. Simi-
larly the conversion is im-
proved by distributing the
thorium within the fuel
elements as homogeneously
as possible, which is achiev-
ed by the choice of very
small coated particles. -
For the other thorium cycles
the kernel diameter has been
chosen of a medium size be-
cause of the maximum of
technological experience
for this type. - The low
enriched cycles use large
19
kernels and shell ball design in order to achieve a high neutron resonance
escape probability by a maximum lumping.
The basis for the fuel cycle costs calculations is given in Table VIII.
2.3 Characteristics of the Equilibrium Cycles
Each of the cases in Table V is analyzed by the methods of Section 1.5.2,
using the V.S.O.P. computer program. The initial enrichment and average re-
sidence times were iterated until a nearly critical equilibrium core of
approximately the desired burn-up was obtained. The results are summarized
in Tables IX through XII.
TABLE IX:Résulta for Equilibrium Cores
- Global Data -
Initial enrichment N,, /N„„
KeffAvg. Burnup
Avg. Residence tine(full power)
Conversion ratio
Core avg. enrichment N /
Power ratio peak/average
Specific power
Fast dosls (> 1O5 ev)in balls
%
MWd/t
days
V %
kwth/kg fiss
1O2WMax. fast doaia (>105 ev)
In outer reflector per year 10 /cm a
Xe override (100 — > 40 %)
Fuel cycle costs
A«eff
mills AWhe
1013
9.71
1.0031
100 058
627
0.549
4.96
2.77
3521
4.14
3.41
-0.020
2.33
1113
7.16
1.0029
69 835
439
0.580
4.20
2.58
4049
2.89
2.96
-O.022
2.63
1213
12.35
1.0010
129 959
815
0.521
5.64
3.18
3064
5.31
3.48
-O.O18
2.21
1313
9.96
1.0000
100 154
627
0.555
5.28
2.80
3315
4.12
3.2O
-O.O19
2.35
1433
12.82
1.0027
130 334
608
0.520
6.87
2.40
3469
5.29
3.49
-O.O16
2.24
4011
7.93
1.0017
100 677
907
0.583
3.93
2.24
3O49
5.62
2.22
-0.019
2.05
4021
7.12
1.0003
101 527
1633
0.625
3.42
2.62
1975
5.68
1.49
-0.016
2.08
9022
3.01
1.0025
36 399
1267 .
0.947
2.81
1.71
1037
4.42
1.08
-0.004
2.89
23502
8.93
1.0032
101 358
1115
0.557
4.19
2.89
2368
4.15
1.74
-0.014
Table IX brings a general characteristic of the reactor performance. The
upper four rows are the set of variables which are chosen for getting a
critical core and a desired burn-up. The initial enrichment is the heavy metal
weighted average of the two different enrichments in the two radial zones.
The global figures in the lower part of this table are complemented by more
detailed figures in the following tables.
Table X gives explicitely the balance of the fissile material which is re-
quired for the economical considerations. The break down of the fuel cycle
costs is given in Table XI, and the thermal hydraulics are covered in
20
TABLE XiRaaulta for Equilibrium Cores
- Mass Balance -
Heavy Metal In Core
Fissile Material In Core
0-233
0-235
Pu-239
Pu-241
Charged per full-power Day
Heavy Metal
0-233
0-235
Discharged per Full-power Day
Heavy Metal
0-233
0-235
Pu-239
Pu-241
Fuel Elements
Ret Balance per Full-power Day
0-233
0-235
Pu-239
Pu-241
kg
kg
kgkgkg
kg/a
kg/a
kg/a
kg/a
kg/a
kg/a
kg/d
kg/a
balls/d
kg/a
kg/a
kg/akg/a
1013
17379
-
657
140
55
29.76
-
2.86
26.51
-
0.582
0.215
0.107
2867
-
-2.278
0.214
0.107
11t3
17876
-
577
122
42
42.52
-
3.02
39.32
-
0.779
0.289
0.131
4096
-
-2.237
0.286
0.132
1213
16861
-
725149
63
22.90
2.79
19.63-
0.465
0.168
0.083
2206
-
-2.334
0.167
0.088
1313
17377
- .
701
145
59
29.76
-
2.92
26.51-
0.648
0.227
0.117
2867
-
-2.282
0.226
0.117
1433
12787
-
697
120
48
23.03
-
2.95
19.74
-
0.568
0.178
0.102
2222
-
-2.J72
0.177
0.102
4011
25015
438
543
2
1
29.58
-
2.38
26.37
0.619
0.172
0.0017
0.0011
1982
0*595
-2.243
- •
-
4021
44390
813
701
3
2
29.34
- .
2.11
26.13
0.583
0.122
0.002
0.001
1931
0.574
-1.994
-
-
9022
102730
2849
45
-
-
82.82
2.505
-
79.67
2.243
0.072
-
-
2553
-0,262
0.072
-
- •
23502
30689
-
950
228
89
29.66
-
2.64 .
26.38
.-
0.430
0.189
0.095
2906
-
-2.210
0.189
0.095
Table XII. The local distribution of the thermal loads on the fuel elements
in the different positions in the core is automatically computed for each
design case. Here, a condensed selection of the results has been made, which
is explained in Table XIII,a and given in the Tables XIII,b through XIII,i.
The Tables XIII show that for most of the cases the maximum ball power out-
put takes place in a relatively high position at the edge of the radial re-
TABLE XI:Results for Equilibrium Cores
- Fuel Cycle Costs -
Fuel cycle costs
Break down:
Fabrication
Reprocessing
Heavy metal expenditure
Heavy metal credit
Net heavy metal
Requirement per full power day:
Uranium ore (0.4X tail)
• Separative work
mills/kWhe.
mills/kwhe
»
••
kg/d
SWU/d
1013
2.33
0.58
0.26
1.75
0.26
1.49
881
424
1113
2.63
0.82
0.39
1.75
0.33
1.42
914 .
414
1213
2.21
0.45
0.19
1.78
0.21
1.57
870
435
1313
2.35
0.58
0.26
1.80
0.29
1.51
904
437
1433
2.24
0.45
0.19
1.86
0.26
1.64
919
462
4011
2.05
0.47
0.21
1.80
0.43
1.37
760
451
4021
2.08
0.48
0.21
1.77
0.39
1.38
; 677
402
9022
2.89
0.97
0.42
2.78
, 1.28
1.50
' - •
-
21
TABLE XII:Résulta for Equilibrium Cores
- Heat Transfer -
Max. power per ball
Max. temp, of gas
Max. temp, ball surface
Max. temp, ball center
Max. temp, gradient atinterface matrix/shell
Temp, at location ofmax. temp, gradient
Core pressure drop
He mass flow
kW/ball
°C°C°C
°C/cm
°Cat
kg/sec
1013
4.85
1000
1006
1074
313
810
0.623
785
1113
4.34
1017
1024
1056
278
797
0.605
785
1213
5.85
10O2
1007
1124
362
811
0.636
785
1313
4.87
1218
1225
1240
322
920
0.464
614
1433
5.71
1206
1229
1274
369
928
0.335
614
4011
4.09
997
1008
1172
270
776
0.603
785
4021
2.69
997
1002
1016
171
755
0.359
785
9022
1.79
1017
1054
1158
111
598
0.314
785
23502
2.78
1007
1011
1017
172
733
0.220
793
flector. Here, the flux is highly tilted due to the reflector effect, and
the fissile material content is high due to the higher enrichment in the
outer fuelling zone. The maximum temperature gradient which is also given
in the tables, does not coincide with the maximum power of the balls. This
is due to the fact that the thermal conductivity is decreasing with an in-
crease in the temperature level and in the fast neutron dosis in the balls
/21/.
2.3.1 Variation of Burn-up
The series of cases 1113 - 1013 - 1213 gives the influence of increasing
bum-up like 70,000 - 100,000 - 130,000 MWd/t for the low enriched cycle.
The influence on the conversion rate is not only due to the increasing ab-
sorption in the fission products, which varies like 7.18 - 8.26 - 9.23 %
per source neutron, but it is also due to the increase in.the axial leakage
varying like 4.93 - 5.42 - 5.82 % per source neutron for the three cases.
The presence of the varying amount of fission products and of the secondary
plutonium isotopes in the lower core causes a shifting of the neutron flux
profile towards the upper part of the core, which causes the change in the
leakage. Coupled with the increasing neutron losses is a decrease in the
fertile material absorption rate, and this brings the reduction in the
conversion which is 0.58 - 0.55 - 0.52 (Fig. 8).
Independent from the conversion is the utilization of the bred fissile plu-
tonium. For the three cases the fraction of bred Pu-239 + Pu-241 being burnt
31
in the reactor directly is 84 - 87 - 89 %.
This means that for the high bunv-up case
only 11 % of the fissile plutonium will be
disloaded and has to be reprocessed. This
fact helps to improve the fuel cycle eco-
nomy because of the low plutonium value.
The combination of the increasing plutonium
utilization and the decreasing conversion
rate explains that only a very slight in-
crease in the daily U-235 consumption is
observed towards the higher burn-up
(Table X). Moreover, Table XI shows that
the uranium ore requirement even goes con-
siderably down.
59
58-
57-
56-
55
S3
Si-
\
\
\ Cast
\ \
\ N
Aratio \
70000 100000 IX
2.7
26
25
2A
-Z2
000 IMWD/ri
fiumtp -
FIG. 8; Conversion and fuelcycle costs as function ofburn-up.
The break down of the fuel cycle costs shows that the tendency is mainly
governed by the expenditure for the fabrication and reprocessing, which is
nearly proportional to the reciprocal value of the burn-up (Fig. 8).
The power peaking in the upper part of the core is due to the product of
the fissile material content in the inserted fuel elements and the thermal
neutron flux peaking. Consequently, it is sensitively dependent on the burn-
up and varies like 2.6 - 2.8 - 3.2 for the three cases (Table IX).
Corresponding to the sensitive variation, in the axial power profile there
is a high difference in the local heat transfer characteristics. From jthe
sequence of the three Tables XIH,c-b-d, can be seen that the ball power
in the hot area at the core bottom is decreasing by a factor 1/2 and, con-
sequently, the central fuel temperature is reducing. The increase of the
power in the upper core area, however, brings for the high burn-up case 1213
a marked peaking of the central fuel element temperature at a position of
104 cm below the surface of the pebble-bed. At this position there is no
peaking observed for the low burn-up case 1113. Here, a very flat maximum
is reached at the axial position of 208 cm and the temperature stays approx-
imately constant towards the bottom of the core. This unsteady switching of
the location of the peak fuel temperature in the core is a typical charac-
teristic of parametric design variations. It is not only dependent on the
32
axial power profile but also on the difference between the temperatures of
the inlet and outlet coolant, and also on the design type of the fuel ele-
ments (see Chapter 4.9).
2.3.2 Variation of the Gas Outlet Temperature
The comparison of the cases 1O13 and 1313 gives the effect of an increase
of the gas outlet temperature on the reactor performance. The coolant flow
has been reduced so that the gas heating rises from 985 to 1190 °C, which
is the highest admissable value with respect to the conservative limiting
figure of 1250 °C for the nominal fuel temperature in Table III.
There is a slight change in the resonance broadening of U-238 and in th«
thermal neutron spectrum, as discussed in /2/ (Chapter 5.6). Herefrom re-
sults a slight increase in the required initial enrichment, in the conver-
sion rate, and in the fissile inventory being somewhat more pregant. The
effect on the fuel cycle costs is very small, and it is mainly due to an
increase of 6 % in the working capital costs for the fuel.
More significant are the changes in the temperature loads. For case 1313
the maximum central fuel temperature is 1240 °C and it is reached at the
bottom of the core. Here, the difference between the temperature of the
surrounding gas and the center of the balls is no more than 22 °C.
This brings some advantage for the consideration of engineering factors
involved with the local power distribution. An uncertainty of 15 % in the
calculated local power density affects the central fuel temperature only
by 3.3 °C. For case 1013, however, the maximum fuel temperature is reached
in' a relatively high position in the core where the power is high. There-
fore, the same uncertainty factor brings an uncertainty of 52 °C for the
figure of the maximum fuel temperature given in Table XII. From this example
results that for the temperature safety margin even a rough estimate re- >
quires the detailed overall core display of the performance data.
For both cases the maximum fast neutron dbsis in the reflector graphite is
appro imately the same, and it will be reached in a position of 63 cm be-
low the surface (Tables XIII,b,e). Here, the gas temperature for the two
33
cases is 507 and 581 °C, respectively. From Fig. 27 (see below) results
that the allowable fast dosis (curve II) varies by 25 - 30 % for these two
temperatures. The allowable lifetime of the reflector varies by the same
ratio! A more detailed discussion of the temperature dosis function for the
reflector will be given in Chapter 4.1i and in /3/ and /4/.
2.3.3 The High Performance Variant
For the case 1433 the gas outlet temperature has been increased to 1190 °C,
the average burn-up to 130,000 MWd/t, the average power density to 12 MW/m ,
and the core height has been reduced by 1/4. The aim of presenting this case
is to outline the developmental potential of this type reactor from the view
of fuel element loads. A comparison between the data of this case and the
permissible ones in Table III shows that it just fulfills the requirements
of the present technology for the fuel elements.
Compared to case 1213 the inventory of the heavy metal is reduced proportion-
ally to the reduction in the core volume. The inventory of the fissile mater-
ial, however, is reduced only by 8 %. A comparison of the Tables XIII, d and
f, shows that for the case 1433 the axial power profile is flatter than for
1213. Therefore, the fuel depletion is more equally distributed from the top
to the bottom, as seen from the axial U-235 distribution. And this brings
an increase in the average concentration. Further, the higher temperature
requires a slightly increased fissile inventory as discussed in the previous
chapter. Adding these effects together, the average N / N ™ ratio is higherfiSS HM
by a factor 1.22 for the high performance variant compared to case 1213.
In the fuel cycle costs the contribution of fabrication and reprocessing is
the same for these two cases. A very slight increase is given for the total
sum, which is due to the higher requirement of ore and separative work. This
does not quite outweightthe reduction in the working capital costs due to
the lower inventory and the slightly better revenue .for the discharged fuel.
An important result is that in the range of this investigation an increase
in the power density does not bring an appreciable improvement in the fuel
cycle costs. It has to be checked whether or not there is an effect on the
total plant costs from the decrease in the core volume. In case that a trans-
34
ition to high power density is not seriously required, one should restrict
to 9 MW/m or even
the safety margin.
to 9 MW/m or even lower because this brings an important improvement of
2.3.4 Low Power Density :
The last of the variants for the low enriched cycle is the previously cal-
culated case 23502 /5/. Compared with 1013 the power density is reduced to
5 MW/m and, correspondingly, the diameter of the core is 12.36 m. As a con-
sequence the leakage rate of neutrons is slightly reduced (7.8 > 7.5 %
per source neutron), and some further slight changes are observed in the
neutronic balance. The only figure which is markedly changed is the fissile
isotopic content. Table IX shows that the average Nf4 S S/NH M ratio is sub-
stantially lower, and in Table X the daily rates of supply and discharge are
different.
unfortunately, the influence on the fuel cycle costs cannot be discussed
here, because the calculation of the case 23502 was made with different
cost data assumptions. The fabrication and reprocessing must be expected to
be essentially the same, because the number of balls being submitted per
day is approximately the same. The fissile material supply is somewhat lower
for case 23502, and the fissile inventory is higher. Therefore no pregnant
tendency can be expected for the contribution of the fissile material in
the fuel cycle costs, when goingto a lower power density and a larger core
design. The same holds for the thorium fuel cycle as will be explicitely
shown in Chapter 2.3.6.
2.3.5 Low Enriched Uranium Versus High Enriched Uranium/Thorium Fuel Cycle
The confrontation of these two fuel cycles is given by the two cases 1013
and 4011. Both cases are approximately optimized fuel cycle variants which
have the same burn-up, the same core configuration, and the same tempera-
ture difference between the inlet and outlet helium. They are different
in the design of the fuel elements and in the N_/N_w moderation ratio.C HM
Primarily the difference between these two cycles is due to the difference
35
in the resonance integrals of the two most important fertile isotopes, and
to the difference in the neutronic properties of the bred fissile isotopes.
The high resonance integral of the U-238 imposes the requirement of a rela-
tively high heterogeneity and a high moderation ratio N /N . The thorium
cycle allows a decrease in the N_/N ratio. Assuming a constant burn-up
this brings a reduction in the daily supply rate of the balls (see increas-
ing residence time in Table IX), and by this it brings a bonus in the fuel
cycle costs by a decreasing contribution of ball pressing and head end (Table
XI). Automatically, an increase in the heavy metal density involves an in-
crease in the fissile material inventory (Table X), which increases the work-
ing capital costs. Further, the fast neutron dosis in the fuel elements in-
creases considerably (Table IX).
The bred fissile isotopes are extremely different in their thermal cross
sections, in their <*- = 0 /fi. and'^=v6'/6' values, and in the equivalence
value for their prices compared to the highly enriched U-235 (Table VIII).
In case 1013 the average thermal« of the fissile plutonium isotopes isa.
in the range 840 - 1310 barn, in case 4011 that of the U-233 is in the range
238 - 253 barn. Consequently, in. 1013 the fissile plutonium builds up to
an equilibrium concentration which is relatively low, and 87 % of the bred
one is directly burnt in the core. This must be considered as an important
advantage as long as a commercial refabrication line is not established.
In case 4011 the U-233 builds slowly up to a considerably higher level, and
30 % of it is contained in the disloading charges. In this study it is
assumed to be reprocessed and sold at the equivalence value of 1.25. This
gives the most important bonus for the Th/U5 fuel cycle costs in Table XI.
Considering the neutron losses in the intermediate heavy metal isotopes
(Pa-233, Np-239), in the fission products, graphite, and leakage, Table XIV
indicates an increase by 3.22 % per source neutron for the case 4011. Here
the loss in the Pa-233 should even be counted twice because it brings simult-
aneously a loss of neutrons and of U-233 build-up. This disadvantage is
compensated by the fact that the Th/U5 cycle needs a lower fraction of neu-
trons per source neutron to be absorbed by the fissile isotopes in order
to sustain criticality. The reciprocal of this figure is the global i value
for the fissile isotopes, which is i\ = 1.91 and 2.07 for the LEU and the
Th/U5 cycle, respectively.
36
The fractional absorption in the fertile material varies only slightly.
Nevertheless, the conversion rate is by 6 % higher for the Th/U5 cycle due
to the higher value of if).
A marked difference between the two cycles is given in the axial profile
of the power density as shown by the peak/average ratio in Table IX, and
by the Tables XIII, b and g. The higher tilt of the axial power density
profile of the LEU cycle is to be explained:.
1. by the higher tilt in the fissile material concentration due to the
lower conversion, and
2. by a higher tilt in the thermal neutron flux distribution being caused
by the presence of the highly absorbing Pu-240 isotope in the lower core.
The resulting effect for the thermal fuel element loads is immediately given
by the display in the Tables XIII, b and g.
For further details of the fuel cycle comparison reference is made to Chap-
ter 3.2 and to /3/ (Chapter 3.1).
2.3.6 Low Power Density
The reduction of the power density from 9 to 5 MW/m for the Th/U5 cycle can
be discussed by the comparison of the cases 4011 and 4021. In order to keep
the total power output constant, the diameter and the height of the core
have been increased from 9.22 —>• 9.96 and 5.00 — > 6.OO m, respectively.
The average burnup and fuel element design have been kept constant. The
case 4021 is used as the reference design for the engineering study.
The corresponding change in the nuclear performance is characterized by ;
the increase in the conversion ratio 0.583 —•» O.625. This is due to se-
veral reasons: A decrease in the leakage 7.12 — ^ 6.52 % per source neutron
brings an increase in the conversion rate by4 C = 0.020. Herefrom results
a reduction in the U-235 feed requirement. Further, in case 4021 the U-233
reaches the equilibrium concentration at a relatively early stage of expo-
sure (Tables XIII, g and h) being due to the increase in the fuel element
residence time. Both facts improve the utilization of the U-233 as seen
from the increase in the contribution to the total fission rate from 41.60
— > 46.83 %. As a consequence there is a change in the global 71 value of
37
the fissile isotopes from 2.073 > 2.095, and this brings a further im-
provement of the conversion rate of^CR = 0.021.
The net effect on the fissile material balance is given in Table X. The dai-
ly supply rate of U-235 is reduced by 11.3 %. Nevertheless, the bonus in the
fuel cycle costs is almost negligible, because the lower expense for the
fissile material is compensated by an increase in the working capital costs
for the higher fissile inventory. Table XI shows that the net heavy metal
expenditure is reduced only by 1.7 %; the credit for the disloaded fissile
material is also reduced because of the lower rates of discharging (Table X).
Consequently, from the view of the fuel cycle economy there is not really
a stringent necessity for an increase of the power density towards the li-
mits of the technological feasibility.
Further, some technical aspects are coupled to the increase of the core heigth
from 5 • • "•> 6 m. First the flow pattern of the balls becomes more uniform.
Second the length of the control rods must be increased, which will be pressed
into the pebble-bed for shutting down the reactor. Third the coolant pressure
drop changes. Fourth the axial profile of the power density becomes sub-
stantially more tilted. Thus the power peak / average ratio (Table IX) is
considerably different for the two cases.
In both cases the conventional fuel element design was used. Therefore, for
case 4011 the maximum fuel temperature is reached in the upper part of the
core. In the 5 MW/m design 4021 the fuel temperature peak in the upper area
is reduced (although the power peaking is higher), and it happens to be exact-
ly the same as at the core bottom. Here, it is no more than 25 °C higher than
the average outlet temperature of the helium.
2.3.7 High Conversion Rate
Case 9022 differs from 4021 by a substantial increase in the heavy metal
content of the balls, a decrease in the burn-up, and by the use of pure
U-233 as feed fuel. These variations allow to achieve a conversion rate as
high as 0.95. It has to be mentioned that the volumetric filling fraction
of the coated particles in the matrix (Table VII) is between the presently
achievable and the envisaged figures in Table III. A suitable way to get
38
this figure down to 0.2 could for instance be a reduction of the thickness
of the outer protective shell of the balls from 0.5 to 0.3 cm.
Table XIII,i shows that this reactor has no more the typical asymmetric power
profile. Consequently the gas is heated more uniformly in the core, the fuel
reaches its maximum temperature at the bottom, and here this maximum is by
104 °C higher than the temperature of the surrounding helium.
Corresponding to the high heavy metal content this fuel cycle requires a
relatively high critical mass (Table X) . As a consequence in the fuel cycle
costs there is a contribution as high as 1.25 mills/kWhe given by the work-
ing capital costs. Another penalty comes from the increasing contribution
of the fabrication, and reprocessing costs being due to the reduced burnup
and to the increased daily throughput of fuel elements.
The physical explanation of the high conversion is given by Table XIV. 44.95 %
of the source neutrons are used to sustain the chain reactions, 12.38 % are
unavoidable losses, and 42.67 % of the source neutrons are available for the
absorption in the fertile materials. The ratio between the absorptions in
the fissile and fertile materials give approximately the conversion rate.
The exact figure, however, is calculated rigorously from the daily product-
ion and destruction rates of the fissile isotopes including the Pa-233 decay
in the disloaded charges.
TABLE XIV:Neutronic Balance
Fractions of fissions in
0-233
0-235
Pu-239
pu-241
Conversion rate
%
*
Fraction losses per source neutron
Absorption heavy metal
fissile isotopes
fertile isotopes
Pa-233, Np-239
fission products
Xe-135
Graphite
Neutron leakage
*
t
%
%
*
%
%
•
1013LEU
0 . 0
52.78
36.62
10.14
0.549
82.53
52.26
29.07
0.10
6.20
2.16
1.44
7.77
4011Th/U5
41.6O
56.85
0.99
0.37
0.583
80.92
48.23
29.88
1.66
10.01
2.19
1.94
7.12
9022Th/03
98.70
0.93
0 . 0
0 . 0
0.947
89.52
44.95
42.67
0.89
5.20
1.41
1.00
4.26
39
Another (private) study yields that a reduction of the burn-up to 15,000
MWd/t reduces the unavoidable neutron losses so far that a conversion rate
of 1.00 is achieved. Here, the fuel cycle costs are by a factor 1.75 higher
than in the case 9022.
In the present study we assume that pure U-233 is inserted as feed fuel,
which is an optimistic idealization. More realistically, the recycling of all
bred uranium isotopes must be considered, a loss at reprocessing of about
0.5 % and a lag time for the Pa-233 decay must be taken into account, high-
ly enriched U-235 has to be used as make-up, and a partial separation of
the enriched U-236 and of the bred plutonium isotopes from the fuel cycle
is required. The result of a preliminary research on the approach to rea-
listic fuel cycles is that an optimal high converter cycle will have a con-
version rate between 0.91 and 0.95, and the fuel cycle costs are expected
to be by a factor between 1.2 and 1.6 higher than for the thorium reference
case 4021 of this study.
2.4 Sensitivity Studies
Two sensitivity studies were performed; one in which the reactor parameters
affecting heat transfer were varied; and one in which the input data for
the cost analysis was varied.
2.4.1 Heat Transfer Variations
In this study, the base was taken to be case 4021. It was assumed that all
nuclear parameters (e.g., power distributions) remained unchanged, while
the reactor power and coolant temperature rise were varied. The results are
shown in Tables XV and XVI. It can be seen that the major result of raising
reactor power and lowering reactor 4 T is to greatly increase the required
helium flow and resulting core pressure drop. The effect of reactor power
on the central temperature of the balls is shown in Fig. 9.
40
i»o2lwa! 3WP MWth
R A O . Ï M T V . •••% • \
MÊSWK- .«.0,-133.«ROM HriOHT /i 0.0 2.19 /2 25.0 Z.Zl I>• '5S.Ô •• . - ^ 3 O .4- ,t5'.O . 2.245 100.0 2.076 125.0 1.347 150.0 . 1-.6?8 .175.0 1.419 200.0 1.21
10 225.0 1.0211 250.0 0.6»12 27&.Q ai 7513 300.0 Ö.ä,414 32Ü.0 0.95
115 350.0 0.4716 375.0- Q„4017 400.0 Ö.3&
118 425.0 '• 0.2919 450.0 0.2520 475.0 0.22'21 500.0 0.1?:22 5Z5-0 0.16,23 55Û.0 O. l i24 .5.75.0 0.14;25 600.0 0. 12
i • . - . • . ' ^ ',
Mo21-b: USpo MWth
iRAO.lNTV. 1 \MESHP. 0.0 133.0)ROW MFI f iHT ' • -
1 0.0 3.29 /2 . 25.0 3.313 iS0;0. • 3.-454 75.0 3.355 100.0 3.106 125.0 2.767 150.0 2.438 175.0 2 . U9 200.0 1.81
i0 225.0 1.5311 250.0 1.3112 275.0 1.1313 300.0 ; 0.9614 325.0 0.8215 350.0 0.7016.375.0 0.6017 400.0 0.5118 425.0 0.4419 450.0 0.3820 475.0 0.3321 500.0 0..2822 5ZS.0 0.2523 550i"0 0.2224 575.0 0.2125 600.0 0.19
Ho21-c: 6ooo MWth
RAD.INTV.Î l " - \MESHP. 1;0.0 -.133 .O\ROW HEIGHT
1 M).O2 25.03 50.0
, 5 100.06 125.07 150.0B 175.09 20O.0
10 225.011 250.012 275.013 3Üd.O14 325.0
4.38 /: 4.41, 4.60î;:4.4'7
4.133.68
• 3.24' 2.ai
2.412.051.75
, 1.501.261.10-
15 350.0 0 .9416 375.0 S 0 .8017 400.018 425.019 450.020 475.021 500.0 '22 525.023 550.024 575.0
0.690.59o.si0.440.300.330.29
' 0.27
Table XV:
,'ÂM.L) /
Power Density Variation at
7 - • } • • • • • • . . . . ,
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41
TABLE XVI:
Variation of Reactor Power Output
Case
Reactor power
Fuel cycle
Gas temperature rise
Max-, temperature of gas
Max. temperatureball surface
Max. temperatureball center '
Max. temperature gradientat interface matrix/shell
Temperature at location of
max. temperature gradient
Core.pressure drop
He mass flow
MWth
°C
°C
°C
°C
°C/cm
°C
at
kg/sec
4021
3O0O
High enriched
2S0 •» 985
997
1002
1016
171
755
0.359
785
4021-a
3000
uranium /
250 -
809
814
886
186
653
0.518
1050
4021-b
4500
thorium
—» 800
809
814
1030
255
701
1.080
1575
4021-c
6000
809
814
1185
342
747
1.827
2099
• • 800 "C gas out lot tamp.conventional bal. ThU. q-= S &%£ 3000 MW,
Ctntarling , « 6ooo (« t ht »5oo MWth
at 3ooo HWth
BOO Icml
FIG. 9: Temperature profile under varyingpower density.
42
2.4.2 Cost Parameter Variations
In Table VIII most of the cost data assumptions correspond to the current
knowledge. A,certain range of uncertainty, however, is given for the data
of fabrication and reprocessing. In order to estimate the corresponding range
of uncertainty in the fuel cycle costs, a sensitivity study has been per-
formed for the sequence of cases 1113 - 1013 - 1213 and for the thorium ca-
ses 4021 and 9022. The cases of the low enriched cycle are different in the
burn-up so that the contribution of the fabrication and reprocessing to the
fuel cycle costs is markedly different.
TABLE XVII:Sensitivity of fuel cycle costs on cost assumptions
Case
Burnup MWd/t
20 % change In cost assumptions for:
Fabrication of coated particles
Manufacturing of balls
Head-end, shipping
Reprocessing
1113
70000
brings
0.064
O.1O1
0.056
O.O24
1013
1OOOOO
1213
13O0OO
4021
1OOOOO
following changes (mills/kWhe)
0.045
0.071
• 0.038
0.016
0.035
0.055
0.028
0.012
0.046
O.O5O
0.025
0.015
9022
36O00
0.131
0.063
0.038
0.047
The results are presented in Table XVII. For the fabrication the highest
sensitivity holds for the dependency on the manufacturing costs of the balls.
In the reprocessing performance the highest sensitivity is for the head-end.
This is a typical mark of the low enriched cycle because of the high modera-
tion ratio, which requires that 101 balls must be manufactured and repro-
cessed in order to handle 1 kg of heavy metal.
For the thorium cycle, especially for the high conversion variant, the sen-
sitivities are expected to be somewhat differently distributed. The rela-
tive contribution of the handling of carbon is essentially lower, which is
due to the lower moderation ratio.
2.5 Consideration of Hot Spot Analysis
The true temperature distribution in an OTTO core differs from the design
temperatures. This difference is due to systematic and statistical devia-
43
tions which result from deviations from nominal values of material proper-
ties, from uncertainties in the determination of these properties, from sim-
plifications in the calculational models, and from the statistical varia-
tions inherent to the pebble-bed.
Previously in this study (chapter 2.3.2) it has been shown that the fuel
temperature at the center of the balls reaches the maximum for many cases
in the upper third of the core, where the power is relatively high. For some
cases it is reached at the core bottom where the power density is strongly
reduced. Thus, an uncertainty in the local power has a quite different res-
ponse in terms of uncertainty in the fuel temperature.
For example for case 1013, at the location of maximum fuel temperature, the
power is 3.03 kW/ball (Table XIII,b). An uncertainty of 20 % in the local
power production would bring an uncertainty of 65.4 °C for the temperature
at the center of the ball. Case 1313 is the same design with the only ex-
ception, that the average gas outlet temperature has been raised to 1190 °C
(Table XIII,e). Here, the maximum fuel temperature is reached at the bottom
where the power is only 0.20 kW/ball, and a 20 % uncertainty in the local
power brings for the central fuel element temperature an uncertainty of
4.4 C only! For both cases must be expected that an overall core hot spot
analysis will yield quite different locations for the expected hottest fuel
elements, and also quite different sensitivities.
Another thing is the uncertainty in the design parameters. This is expected
to be different for different design of the fuel elements, depending on the
material data and on the methods of manufacturing. For a final concept they
will be specified by the company who will offer, produce and guarantee them.
The conclusion is that the performance of a binding hot spot analysis is not
advisable and also not possible for a conceptional survey study, as given
here. A previous unpublished research on the optimized design case being de-
rived for a helium turbine plant /4/, has given an uncertainty of about
100 °C for the maximum fuel temperature. This figure is well covered by the
safety margin of 2O0 °C, which is already incorporated in the conservative
restriction of 1250 °C for the nominal maximum fuel temperature in Table III.
44
CHANGING-OVER IN FUEL CYCLE
3.1 Introduction i
Detailed research on the low enriched fuel cycle, thorium cycle, thorium-
Ü-233 recycling cycle, and on the high converter is given in the references
/2/, /3/, and /16/, respectively. For these four cycles a reoptimization
has been performed assuming the core design of the THTR. And the switch-over
between these cycles has been investigated. '
Compared to the THTR the thermal power output has been raised from 750 to
1000 MWth, and the gas outlet temperature from 75O to 1050 °C. This choice
fits well into the range of the present test experience for the fuel elements
and fulfills the requirements of process heat application. It turns out that
the reactor can be operated in all three of these fuel cycles and that the
change-over between them can easily be performed /17/.
3.2 Different Fuel Cycles
An optimal low enrichment fuel cycle requires a relatively high resonance
escape probability which is achieved by a N r/NH M ratio as high as 355 (Table
XVIII). First, this minimizes the build-up of the parasitic absorber Pu-240
in the long exposed fuel at the lower core which would cause a too high power
peak in the fresh fuel at the top. And second, it helps to minimize the en-
richment requirement for the feed U-235. By the insertion of 10.6 % enriched
! TABLE XVIII. • Design datadooo MWth>
: . .
Average power density MW /n
Height . cmRadii of 2 fuelling zones cmGas temperature inlettoutlet CRadii of shell ball cm
Fuel cycle
diameter of coat-part-kernels ,/tThickness of 2 coatings fju
Fuelling zonesModeration ratio N-/N™Fract. of dummy ballsVolume fract. coat, part in matrix -Heavy metal density in matrix gr/cmEnrichment N-. /NUM \(Recycling) f l s s "" %
8
H672o6 / 2922 So -. loSo
l.S / 2.S / 3.o
Lou enrich-ment
8oollo/ 8o
1 2380 32oo.o o.oO.o74 o.o87O.193 0.2289.IS 12.58
Thorium
Soo80/80
1 • • 2 •• ••
28o 230O.176 0.0O.133 O.133O.3o8 o.3o88.O3 9.88<7.So)(9.33>
Hi.ConversionC Z " U-feed)
•tooSo/80
1 2llo llo0.0 ' 0.0
6.3UO o.34oO.632 O.6322.92 3.S2
45
uranium, a burnup of 115,OCX) MWd/t is. achieved. The conversion ratio is 0.56,
and it turns out that 86 % of the bred fissile plutonium is burnt in the
reactor itself. The fuel cycle cost amount to 2.12 mills/kWhe. This indicates
the low enrichment cycle to be a reasonable choice as long as the remote fuel
element manufacturing is being developed for the alternative thorium cycle,
or if the supply of high enriched uranium is problematical for any non-scien-
tific reason.
In the fuel cycle of high enriched uranium and thorium, the heavy metal load
per fuel element can be increased without penalty in the fuel balance or in
the technical performance. This helps to decrease the contribution of fa-
brication costs in the economical balance, and further, it helps to increase
the breeding of valuable U-233. For a Nç/N^ ratio of 260, the fuel costs
turn out to be below that of the low enrichment variant assuming the cost
data of Table XIX.
Table XIX
... Cost data assumptions
U308Conversion U3O8»UFSSeparative workTail enrichmentPu-fissileU-233ThoriumCoated particles fabricationBall fabricationReprocessing-Ilead end * transportReprocessing-HMLead and lag tinesAnnual load factorNet electrical efficiencyInterest rateTax rate on fiss. mat.
103.M100.0036000
18000102203.tu
10. U9 8 . "U
219-3600.8O.U10»2%
S/lb2/kg U met.S/kg SWU
S/kR2/kR*/kB
ï/kg HM2/ball
. S/kf» Graphitel/kg, HMd :
As soon as remote fabrication is readily developed, one may follow the feed-
breed concept. A promising variant of this one is to insert a mixture of three
different types of balls : . . . .
Type 1: containing the bulk fertile material,
Type 2: containing the make-up fuel, • .
Type 3: containing coated particles with a mixture of the
recycled uranium and thorium.
This concept requires remote fabrication only for the last type, which re-
presents about 1/3 of the daily supply rate of balls. The three types can
be marked by engravings, which allows to separate the recycle type from the
others in the discharge batches.
46
The fourth fuel cycle variant is the high converter. It will become an
attractive concept as soon as shortage in fissile raw material comes up.
This cycle will reach its ultimate perfection in partnership with fast breed-
ers producing U-233 in the blanket. The conditions for raising the conver-
sion up to 0.90 are:
Manufacturing of balls with a N_/N„„ ratio of 110, which is present-er HMly the goal of development,
Reduction of the burn-up to 28,O00 MWd/t,
The use of U-233 as feed fuel.
An increase of the conversion up to 0.96 can be achieved by three actions: .
Operating the reactor at base power reduces the short term control require-
ment. The transition to large scale plants of 3000 MWth reduces the neutron
leakage losses. A reduction of the power density to 5 MW/m decreases the
neutron losses in protactinium and xenon.
For two reasons the pebble-bed reactor is favorite for high conversion:
First, the continuous fuelling performance does not require excess react-
ivity for long term control. Second, it allows to vary the conversion by
TABLE XX:The four different equilibrium cycles
looo MWth, 8 MW/nT , 25o lo5o°C
Fuel cycle
N_/NHM in feed batches
Av. fuel residence time days
Average burn up MWd/t
Conversion ratio
Power peaking
Control rods withdrawal Ak/k
Xe-Override loo - 4o% Ak/k
Fissile inventory kg
supply gr/d
U-nat.' requirement . • kg/d
Separative work kg-SWU/d
Fuel cycle costs mills/kWhe
Fabrication only mills/kWhe
Low enr.
355
848
115 ooo
o.S4
2.8
o.o2o
-0.0I6
345
916
217
I60
2.12
O.49
Thorium
259
Io94
114 000
O.55
2.6
0.0I8
-0.0I9
361
783
19o
168 '
1.92
o.4l
Th-recycle
259
1113
115 000
0.60
2.5
0.0I8
-0.0I8
36o
729
135
12o
1.84
O.41
Hi-conv.
llo
63o
28 000
o.9o
1.3
o.oo5
-O.OO7
62o
Io82
2.91
1.2o
47
continuous changes in the fuel residence time, i.e. in burnup and thus to
adapt to the requirements of the market.
The control of the reactor is realized by moving rods in the cavity above
the surface of the pebble-bed. Table XX shows that the efficiency of a cer-
tain number of control rods is substantially different for the three cycles.
But the reactivity requirement for the load follow varies approximately in
parallel. This fact allows the utility which operates the reactor to change
over from one fuel cycle to another.
3.3 Switching-over between them
For the four discussed cycles a reactor life sequence has been explicitely
followed which involves the start-up, running-in and switch-over between
them. The first core loading starts with balls of 7 different enrichments.
From the bottom to the top they are arranged gradedly, so that the typical
asymmetric power profile is almost obtained at the start of life. This allows
to achieve a short and smooth running-in period. .
In Fig. 10 the set of
curves in the middle pic-
ture gives the power load
history which the balls
pass through when they
are moving through the
core at the center line
from the top to the
bottom. The figure shows
that the running-in pe-
riod can be considered
to be essentially fi-
nished after about 7
months, which is 1/4 of
the mean residence time
of the fuel elements.
SWITCH-OVER IN FUEL CYCLES
14 16 TEARSTEUCHEItT-HMS.KFfl
FIGURE 10:
Switch-over in fuel cycles
The switch-over to the thorium cycle shows a variation in the maximum of
the local power for a time interval of about 4 months. During this period
48
an intermediate layer of fuel elements has been inserted, in which the fiss-
ile and fertile material content is somewhat lower than in the equilibrium
charges of the subsequent thorium cycle. By this way it is possible to avoid
an undesired power peaking at the interface between the charges of the diff-
erent cycles. The start of recycling of the bred U-233 indicates nearly no
effect in the fuel element loads. '
The last section of this reactor life display shows the switch-over to the
high conversion cycle. Here, the power profile immediately flattens to the
shape of the new equilibrium. This is no more the typical asymmetric OTTO
curve. At high conversion, the fissile material concentration in the fuel
elements changes only slightly during exposure. Therefore, the axial dis-
tribution of the power density is almost flat, and the fuel temperature at
the ball centers increases by 110 °C for the balls at the core bottom.
For the whole history the k curve shows a swing of less than 1 % in react-
ivity, which can be covered by thé current control system. It has been brought
into this flat shape by changes in the flow velocity of the balls, which is
shown in the lower drawing of Fig. 10. The possibility of varying the daily rate
of fuel element supply is a typical mark of the continuous onload fuelling.
It represents a very useful degree of freedom when operating this type of
reactor.
3.4 Mass Balances
LOH CNKICHPICNT
IGR/DBY
400
200-
K0
HILLS/KHHEi lOTfiL
•FISSILE.
FISSILE ET
THORIUM CYCLE
FISSILE HATER
U-2»
FISSILE INVENTOTHL
FUEL CYCLE CO
-FUBBICBTIOM ,t -PROCESSIMO
TH-RECTCLINO
fiL BALANCES.SUPPLY0-OISCHRRCE
u-tn is-oi
ORY
iTS
HI-CONVERSION Corresponding to the supply
of b a l l s , the supply of feed
fuel i s strongly varying dur-
ing the periods of running-in
and cycle changes. In Fig. 11
the daily supply rate i s g i -
ven at the upper drawing. At
the beginning of the low en-
FIG. 11:Mass and cost balance forsequence of fuel cycles
10 12 14 16 YEARSTEUCMERT-MHHS.KFH
49
richment cycle, the U-235 supply has to counterbalance the build-up of the
plutonium isotopes. This brings the strong variation during the running-in.
Although for the low enriched and thorium cycle the conversion ratio is al-
most the same, the disloading rates of plutonium and U-233 differ by a fact-
or 2. The calculation shows that under the equilibrium operation of the
thorium cycle 25 % of the bred U-233 will be disloaded and must be submitt-
ed to the reprocessing. At the low enrichment cycle only 14 % of the bred
fissile plutonium comes to the disloading and 86 % are burnt in the react-
or directly. This fact results from the great difference between the thermal
cross sections which causes a high difference in the equilibrium concentrat-
ions as shown in the middle drawing.
The total fissile inventory which is also given in the middle drawing starts
for the initial core by 47 % below the equilibrium value of the low enriched
cycle. It turns out to be comparable in the thorium cycle and to increase
for the high converter. This tendency suggests to start a high converter in
any case from a low conversion cycle which requires a minimum first core in-
ventory. By this way the high level of fissile isotopes, being required for
the high conversion, will partly be built up in the core by itself.
3.5 Fuel Cycle Economy
Table XXI gives a break-down of the fuel cycle costs, and the lower draw-
ing in Fig. 11 gives the variation during the fuel cycle changes. The curve
of the net fissile costs gives the difference between the expenditure and
TABLE XXI:
Break down of Fuel Cycle Costs (raills/kWhe)
Total, equilibrium cycle
Fissile material
Fiss. mat. net
Fabrication
Reprocessing
Lowenriched
2.12
1.60
1.U1
o.U9
O.22
Thorium
1.92
1.67
1.32
0.41
. O.18
ThoriumRecycle
1,84
1.6o
1.25
O.41
0.17
HighConverter
2.93
2.79
1.16
1.20
O.52
50
revenue for the' heavy metal and fissile material including the working ca-
pital costs. It changes only slightly for all discussed fuel cycles. The
more sensitive changes in the total fuel cycle costs are due to the contrib-
utions of the fabrication and reprocessing. For the high converter this con-
tribution is as high as 60 % which results from thé low burnup, i.e. from
the high through-put of fuel elements.
It appears that under present conditions the thorium recycle is the most '
attractive concept provided that the realization of refabrication will not
bring a substantially extra cost penalty.
TABLE XXII:
Fuel Cycle Costs (mills/kWhe) under Cost Data Variations
Reference cases
Fabrication coat. part. 22o •» llo 2/kgHM
Fissile material costs multiplied by 1.2
Interest rate lo% • 15t
Assumptions for the future
Lowenriched
2.12
2.06
2.41
2.21«
3.06
Thorium
1.92
1.82
2.19
2.10 .
2.69
ThoriumRecycle
i.8t
1.71»
2.O9
. 2.ol
2.54
HighConverter
2.93
2.52
3.17
3.26
2.71«
In order to see if these relations hold also under changed assumptions some
basic cost data have been varied such as coated particles fabrication costs,
fissile material costs, and interest rate (Table XXII). The reduction of
coated particles fabrication costs by 50 % brings a slight advantage to the
thorium and thorium recycle concept similarly as the 20 % increase in fissile
material price does. The low enriched cycle would hardly become completely
competitive even by increasing the credit for the discharged plutonium from
6 to 15 #/gr Pu-fiss. The increase in the interest rate from 10 to 15 % brings
a slight but not decisive improvement for the low enriched cycle.
Since a degression in costs might be expected when increasing the size of
the reprocessing and refabrication facilities> and since an increase in the
fissile material prices in the future seems to be a reality, the last case
can be understood as an extrapolation to the future situation. It assumes
an increase in the fissile material costs by 100 %; the costs for fabrication,
refabrication, and reprocessing are reduced by 50 %; and the interest rate
51
is assumed to be only 5 %. These assumptions cause a substantial change in
the cost relations. The highly converting fuel cycle becomes very attractive.
It becomes even less expensive than under the reference assumptions, where-
as for the other three cases the fuel cycle costs increase strongly.
52
PREVIOUS RESEARCH ON TRENDS
4.1 Introduction •
The first computer run on the idea of the OTTO fuelling scheme has been per-
formed in June, 1970. Since that time a number of survey studies and some
detailed researches have been performed which are listed under references
/I/ to /23/. The aim of this chapter is to outline briefly the items being
discussed up to now. At each item the reference will be given for the de-
tailed explanation.
The basic feature of this, type of reactor is the mutual coupling between the
neutron flux distribution, the movement of the fuel elements, the accumulat-
ing burn-up, and the geometric configuration of the core. Herefrom results
the individual power density distribution (Fig. 4) being responsible for
the thermal hydraulic characteristics. The previous studies mainly concern
the possibilities for influencing these correlations.
Three basic reports have been written up to now: Ref. /2/ brings a survey
study for the low enriched fuel cycle, ref. /3/ the same for the highly en-
riched uranium/thorium cycle, and ref. /4/ gives a detailed research for a
2700 MWth reactor in thorium cycle considered for a helium turbine plant.
The latter one uses a design of core and fuel elements which is very similar
to that one used for case 4011 in chapter 2 of this study. Therefore, all
detailed results obtained in /A/ will hold also for this design variant in
Table IX, in a very good approximation.
4.2 Isotopic Distributions in the Core
For a thorium cycle the depletion of U-235 and the build-up of U-233 is
shown in Fig. 2 (see also ref. /3/, Fig. 14). For the low enrichment cycle
see ref. /2/ (Fig. 2). From a comparison of the two cycles (/3/, Chap. 3.1)
results that the build-up of Pu-240 in the lower core has an important in-
fluence on the power density profile in the core. Fig. 12 shows that the
Pu-240 absorption is in the range of 16 % of the total absorption rate at
the bottom. In order to minimize the power peak at the top it is advisable
53
0.015-
0.010-
o.oos
HRCROSCOPIC CROSSECTIONS
NUwFIS
Low enriched
Th/U5
100 300 500 CM
TOP BOTTOM
FIG. 12;Axial distribution ofmacroscopic cross sections.
to use a N_/N ratio inC HM
the range 350 - 420 for
the low enrichment fuel
cycle.
The distribution of Xe-135
is discussed in /2/ (pg.16)
for different power densi-
AG-110M #)
RG-111)SCALE ENLARGEDBY FACTOR lOO
ties. A broad discussion
of the distribution of the.
different fission pro-
ducts is given in /2/ (Chp.
5.5) and in ref. /7/.
Symptomatical for the diff-
erent possible distributions
of the fission products is
a series of three silver iso-
topes as given in Fig. 13.
Ag-lO9 results from direct
fission yield and from a
chain of Pd-isotopes. Apart
from a slight partial delay
its build-up rate is pro-
portional to the FIMA and the volumetric heavy metal content, which are both
higher in the outer loading zone. The rate of destruction is relatively small.
It is only given by neutron absorptions, which yield with the probability
of 3.63 % the Ag-llO m.
BUILD UP
YIELD-»
YIELD-»
YIELD-»
YIELD-»
OF
4.
SILVER
5«106A
STABLE
IN V.S.O
PD-1071PD-108
STABLE
219 D
7.5D
P.:
"XG-109| 3.63Ï
AG-110M1
AG-111
FIG. 13; Local distribution of silver isotopes
Ag-llO m has a half life of 249 days, and in the lower core the rate of de-
cay predominates over the build-up. The slope of decrease is steeper in the
outer radial zones, which is due to the higher residence time of the outer
balls. At the bottom the concentrations are reduced approximately by a factor
of 1/2 compared to the maximum in the middle.
54
Ag-111 results mainly directly from the fission rate, and the low half life
of 7.5 days causes an extreme reduction in the lower core.
The silver isotopes are of special interest because of their high diffusibi-
lity in graphite. In reactor applications with a gas temperature level higher
than 850 °C a considerable rate of release has to be expected. The figure
indicates that a great amount of the unstable silver isotopes decays already
before reaching the hot area at the bottom. •
4.3 First Core and Runnlng-in
Three studies of the reactor start-up have been performed. The first one is
for the low enriched cycle (/2/, pg. 50)'. It points out the feasibility of
a well-balanced and short running-in. In the fresh core balls of different
fuel enrichments are gradedly inserted so that the asymmetric power density
profile of the envisaged equilibrium cycle is almost obtained at the start
of life. Natural uranium is used for the balls at the lower part of the core
in order to reduce the costs. From the beginning the continuous fuelling pro-
vides the same type of fuel elements as for the equilibrium cycle. With re-
gard to variations of the local power peaking the running-in period can be
considered to be finished after 1/3 of the core has been replaced by the
shuffling. The performance is very similar to that one which is given above
in Ch. 3.3.
In order to minimize the disloading of low burn-up fuel, the second study
(/3/, pgi 37) admixes a high fraction of pure graphite balls between the
fuel elements inserted in the lower part of the core. This is possible be-
cause the power density is low in this area, anyway. That study uses the
Th/U5 cycle. It shows further a very smooth transition to the recycling, as
given here in Ch. 3.3.
The application to a 2700 MWth (- gas turbine -) plant in the Th/U5 cycle is
given in ref. /4/, pg. 14. As this design and fuel cycle is very similar to
case 4011 of the present report, we give here some more detailed informa-
tion in the table and figures. In Fig. 15 the saw tooth curve is the k ^ f as cal-
culated by the discretized simulation of the continuous reactor operation, and
the continuous curve gives the corresponding interpolation. The lower picture
55
TABLE XXIII:Fuel Element Identification for the First Core
Brennelenienttyp Nr.
ModerationsVerhältnisVNSM
B l i n d k u g e l a n t e i l *
A n f a n g s a n r e i c h e r u n g %
S p a l t s t o f f pro Kugel g / K u g e l
1 2 3
25o 23o 18o
7 . 3 8 o . o o . o
8 . 4 o 12 .Uo 2.4O
1.35 1 .99 0 . 4 9
4 5 6 7 8 9
25o 23o 25o 23o 23o 12oo
7 . 9 3 o . o 7 . 9 5 o . o o . o 8 o . 5 5
«+. 32 6 . 8 4 2 . 5 9 3 . 8 4 2 . 7 4 2 . 5 9
O.69 1 . 1 ° O.42 O.62 0 . 4 4 o . 4 2
FIRST CORE LOflOING
FIG. 14:First core loading.
gives the interpolation of the power
density at 11 equidistant points on
the reactor axis.
4.4 Influence of the Core Height
The power load history of the fuel ele-
ments is sensitively dependent on the
height of the pebble-bed, which is one
of the most significant distinctions
from other reactors. The height fixes
the distance between the fresh and the
old fuel elements, and i ts variation
influences the
axial neutron
flux distribu-
tion.
200-
400-
600-
TOP REFLECTOR AND VOID
1
4
4
6
6
6
6
6
9
9
1
4
4
6
6
6
6
6'
9
9
1
4
4
4
6
6
6
6
6
9
9
1
4
4
6
6
S
6
6
9
Slu
i
b7
7
7
8
8
8
q
9
9
BOTTOM REFLECTOR
35
~557
778
T
9
9
3DX
,—
mFLE
-4Q3D
RUNNING IN PERIOD
1.10-1.05-1.000.95-
/flVERflCE
FI6. 15:Running-inperiod
20-
15-
10-
5-
flXIflL PflMERlMM/M3)
! UPPER CORE
; LOWER CORE
500 1000 1500 2000 2500 OflYS
56
•aM3
25-
20-
15-
10-
5-
flXIHL POHER PROFILE DEPENDING ON CORE HEIGHT
30.8=TEST EXPERIENCE
CURVE HEIGHT
467 CM427 CM386 CM
^ ^ . nvERRGE
RADIUS
292305321
CMCMCM
100 200TOP
FIG. 16: Axial power densitydepending on core height.
The effect of height va-
riation on the power den-
sity profile is given in
Fig. 16. Curve I holds for
a low enriched design which
is similar to that of
Table XVIII. A brief study
has changed the core di-
mensions under constant
volume and total power
output. The height has
been varied as 4.67 — >
4.27 — > • 3.86 cm, and
correspondingly the ra-
3.21 cm. The set of curves I —>• II r-> III gives
300 400 500 CMBOTTOM
dius 2.92 —*• 3.05 —
the corresponding flattening of the power density profile. In that study
the fuel element residence time and the initial enrichment have been kept
constant. The neutron leakage has been observed to vary by O.O4 % per source
neutron, and kgff between 1.OO3 and 1.OO07.
Briefly can be concluded:
Shortening the core b y A h and keeping the volume constant requires
an increase of the radius. This brings a considerable reduction of
the power peaking in the upper core; and a reduction in the pressure
drop roughly proportional to the cube of h.
The axial power density profile stays almost constant, if the core is
shortened by "cutting off" the lower part under corresponding decrease
of the burnup and increase of the power density. For that case the
pressure drop reduces proportionally to^h.
Further point out of this concern is given.here in Chapter 2.3.6.
4.5 Variational Studies
Variational studies have been performed for the low enriched and for the
thorium cycle. They have preferably been evaluated in terms of technical
aspects, and secondarily with respect to the fuel cycle costs.
57
POWEROENSITT
30-
25-
IS-
10-
HH/M3 . .
IvVin
x
30 ,8 • Fuel element
Identification:
VNHM
Low enr.rh/.U5
23o
III
test
35o
IIIIV
100TOP
300 500 CMBOTTOM
FIG. 17;Power Density ProfileDepending on Fuel Cycle andModeration Ratio
For reducing burnup, the
axial power profile is
flattening (/2/ pg. 18,
and Ch. 2.3.1 of this
study), and the conversion
ratio improves.
The effect of moderation ra-
tio variation is different for the two fuel cycles. Fig. 17 shows that de-
creasing the moderation ratio 350 — > • 230 makes the profile steeper for
the low enriched cycle and flatter for the thorium cycle (/2/ pg. 20, /3/
pg. 15 and 22). The reason is mainly given by the different bred isotopes
in the lower core as explained here in Ch. 2.3.5. .The conversion ratio im-
proves for the thorium cycle and not necessarily for the low enriched one.
The high conversion variant, case 9022, is a synthesis of low burnup and
low moderation ratio. .
The choice of the average power density influences the power profile slight-
ly for the thorium cycle (/3/ pg. 27), and has no effect in the low enriched
one (/2/ pg. 13).
The size of the coated particles and of the ball design has very low effect
for the thorium cycle. For the low enriched cycle the heavy metal should be
arranged as heterogeneous as possible (/2/ pg. 22, /3/ pg. 28).
Recycling of the U-233 in the thorium cycle has only very low effect on the
power distribution, independently whether or not the U-236 is separated (/3/
pg. 29). . •
Changes in the temperature rise of the helium coolant requires slight changes
in the feed enrichment. The effect on the power density profile is very low
(/12/, and here at Ch. 2.3.2).
Insertion of plutonium allows to flatten the power profile as strongly as for
the high conversion variant (Fig. 18).
58
3-
1 -
Normalized
Top
I Low enriched
II Thorium cycle
III Thorium/plutonium feed
IV Low enriched, half height
FIG. 18:Power density pro-file for variousdesigns.
200 400 600 CM
Bottom
4.6 Manipulation of the Radial Power Density Prof i le
Pig. 19 gives two different functions for the power density distribution
in a 2-zone core. For both of them the radial gas outlet temperature pro-
file is very flat, because the radial dependency of the axially integrat-
ed power density is flat (/3/ pg. 31). For the fat curves a higher enrich-
ment is used in the balls of the outer zone. For the function of the thin-
ly drawn curves the enrichment is the same for the two zones, but there is
a 35 % contribution of pure graphite balls admixed in the charges of the
inner zone, so that the moderation ratio is accordingly higher.
Fat curves: 2-zone enrichments.Thin curves: 2-zone moderation ratios
REFLECTORS
BOTTOM
FIG. 19: Power density for varying 2-zone loading
59
It results that a combination of grading the moderation ratio and the enrich-
ment for the two zones allows simultaneously to flatten the radial profiles
of the gas outlet temperature at the bottom and of the maximum ball power in
the upper part of the core. This has been done for the cases of the present
study (cp. Ch. 2.2).
4.7 Fuel Cycle Comparison
The low enriched uranium and thorium fuel cycles have been compared here in
Ch. 2.3.5, Ch. 3.2, in /3/ pg. 15, and in /17/. The main conclusion is that
the low enriched cycle requires high moderation ratio, which brings a slight
penalty for the fuel cycles costs. The closure of the outer cycle is not
necessarily required, which makes it advisable to use' this variant for the
near future concepts.
The thorium cycle is not restricted by such condition, which allows a higher
flexibility for optimizations. The high conversion variant (Ch. 2.3.7, Ch.
3.2, /17/) requires testing of fuel elements with heavy metal loading of
30 - 40 gr per ball, and it requires the closure of the outer fuel cycle.
It is a very promising short term supplement to the fast breeders.
4.8 Ball Flow Distribution
The trajectories of the movement of the balls requires an experimental con-
firmation for each core design. In /2/ pg. 55, an estimation is given for
the uncertainties being involved with the currently used model with respect
to the fuel burnup history. The used method is a variation of the computa-
tional model.
Fig. 20 gives two different ball flow schemes which are supposed to be an
extremely flat and an extremely tilted one. Correspondingly, the right hand
picture gives the axial distribution of the burnup. The maximum uncertainty
at the core bottom is about 4.8 % compared to the mean value. For the maxi-
mum temperature rise of the helium there is a difference of 5 % between these
two cases.
60
l.S
1.0
o.s
(R) n
12/tS/lU/lB
11/13/15/17
CHANNEL I II II! IV
60000
«0000'
20000-
HMD/1
/
/
>
60000
UO000-
20000-
MHD/T
' /
U/13/1S/17
/ ^ ^ ^ 12/13/tH/ie.
200 UDO 800 CH
BOTTOM
100 200 250 CH
FIG. 20: Varying ball flow pattern and correspondingburnup distribution.
For large core designs with more than one disloading tubes the calculation
requires a 3-dimensional burnup code which is presently not available. A
series of different 2-dimensiorial flow schemes have been followed in /22/,
which indicates for the burnup and gas outlet temperature a certain range
of deviation from the mean value, being caused by a given deviation from
the mean value for the ball velocity above the disloading tubes. For one of the
gas turbine design cases of ref. /4/ was found for the gas outlet tempera-
ture an uncertainty of 15 % compared to the temperature difference between
the inlet and the average outlet, assuming a pessimistic flow scheme of the
balls.
4.9 Temperature Distribution in the Core
The strong dependency of the temperature distribution on the power profile
has repeatedly outlined in section 2. The most important features are:
In the hot area at the core bottom the temperature differences in the
fuel elements are very small (Fig. 21).
A 2-zone loading scheme allows to achieve a very flat radial tempera-
ture distribution (Fig. 21).
The central fuel element temperature reaches a relatively high value
already in the upper core area and stays almost constant. Depending
61
FIG. 21:Power and temperature dis-tribution for variation ofgas outlet temperature.
RXIflL PROFILE OF POHERDENSITÏ Q UNO TEMPERflTURES T[HT CENTER OF BflLL. SURFRCE, GBSI
HW/M3
25-
20-
15-
10-
5-
T•CELSIUS
•1500
25D=INLET
1250
- - - ^ S ? 5 ^ — • ' ™ 1150=flV.0U
: £ V ~ " 7S0=BV.0UT
\?><v«^0Bo7H cflSES
200
TOP400BOTTOM
500 CN
RBDIRL PRBF1L OF TEMPERflTUHES fll THE BOTTOM(HT CENTEn OF BflLL, SURFHCE. GflS)
on the power density profile
it reaches the maximum in
the upper or in the lower
area, and the influence of
uncertainties in the local
power production is strong or
or weak, respectively (Fig.
5 and 22).
- An increase of the gas tem-
perature rise /9/ influenc-
es the fuel temperature pro-
file as given in Fig. 21.
Here, the average outlet
temperature is raised from
750 to 1150 °C. The maximum
fuel temperature increases
only by 300 C and shifts
towards the core bottom.
- If the maximum is in the
area of high power produc-
tion, the shell ball de-
sign helps efficiently to
reduce the peaking (Fig. 22).
- The shell ball design re-
duces substantially the
peak temperature, but on-
ly slightly the mean value in the fuel elements (Fig. 23).
- Another instructive display is given in Fig. 24 (cp. /1O/). The function
being plotted is the volume fraction of fuel at given intervals of tem-
perature and FIMA. It is approximately proportional to the duration of
the fuel exposure in these intervals. The upper figure holds for the
THTR-type in which the fuel is circulated several times through the core.
The lower figure holds for the OTTO scheme; it shows a broad flat ridge
which cumulates in a sharp peak. The burn-up proceeds rapidly up to 9 %
•
1500 •
•
•
1000 •
•
soo •
•
CELSIUS
/1250
___
y—-;==^s==:
M150=0U1LET
N750=0UTLET
\2S0-INLET
100 200 300 CMRflOIUS
62
POMER AND TEMPERATURES HT THE AXIS OF THE CORET
CELSIUS1500-
1000'
770-AV.OUTLET
500
0KH/6ALL
S
•<*
A
ÎK
\I «
fts
ffi
s
m
. Q
s.
—«
S *
7 »TECHN.LIMIT 0
iO-T€ ÎHNiL
S—
iMIT t
FIG. 22:Thermal loads for con-ventional and shell ball.
100TOP
200 300 400 S00 CMBOTTOM
'/• A
100
ec
60-
no-
20-
THTR
OTTO
SHELL BfllL
1250 'C=LIMn FORFUEL TEMPEHHTURE
200 UOO 600 800 1000 1200"CELSIUS
FIG. 23; Fraction of fuel above given temperature
FIMA, and afterwards the fuel stays for a relatively long time in the
temperature interval between 740 and 780 °C. For the design case of the
Figures 23 and 24, the average gas outlet temperature was 750 C. The
figures show that this is also about the temperature of the highly irra-
diated fuel, which is important to know with respect to the fission pro-
duct release.
4.10 Thermal Conductivity
The thermal conductivity A is sensitively dependent on the temperature and
fast neutron dosis (cp. /1O/, /3/ pg. 47). All calculations take into account
this 2-parametric function as measured by BINKELE /21/. A typical applica-
tion of this function is given in Fig. 25 (which holds for the case of
Fig. 22). Here, the upper curve holds for the surface and the lower one
for the center. The strong decrease in A during the first half of the inser-
63
FRACTION OF FUEL flT GIVENT H T R TEMPERATURE PND BURN UP
TEUCHERT-HRH5, KFH
FIG. 24:Fraction of fuel at given tem-perature and burn-up.
tion period is predominantly
due to the dependency on the
dosis. The flat shape in the
second half is due to the
assumption of a constant A for
a dosis higher than 3*102
N/cm , which has been made in
default of experimental data.
This, however, is of minor
importance for the OTTO in
which the irradiated balls
produce only low power. Here,
an uncertainty of 30 % in À
results in an uncertainty of
• no more than 2.6 °C for the
central temperature of the
old balls at the bottom of
the core.
In the experiments of BINKELE and HASENCLEVER the fast flux exposure has
been performed at the constant temperature of 935 °C. From the whole set
of their experiments one can draw the conclusion that the A -function might
change by 10 to 3O % if the variation of the sample temperature during the
W/(CM«°C)
. .0.4-
0.3-
0.2-
0.1-
1021 NVT
•8
t THERMALL UPPERVî L0HER
•4%TTT>
2 / /
CONDUCTIVITYCURVE: SURFACECURVE: CENTER
X^^-
FAST DOSIS•
TOP100 200 300 400 500 CM
BOTTOM
FIG. 25: Thermal conductivity in axial positions in the core.
64
exposure would be respected. Therefore, the calculational display given
here might be helpful in setting up experiments which simulate the real
history of the fuel element exposure.
4.11 Fast Neutron Dosis at Reflectors
FBST FLUX.SECI
1.0
o.s
0031S
Top-reflector
Bottom reflector
200 U00 eoo CH
FIG. 26: Fast neutron flux atthe edges of the core.
HV1 FAST DOSIS (E>0.l MEV)
ÎEHJP01HT
10
10500 tooo 1500 C
FIG. 27: Fast dosis vs. temperatureat side reflector in 30 years.
Fig. 26 gives the annual
fast neutron dosis at the
inner edges of the reflect-
ors at the top, bottom, and
side of the pebble-bed, res-
pectively. It holds for
the 1-zone and 2-zone de-
sign cases in /2/ pg. 49,
which have an average power
density of 5 MW/m only.
As soon as the power density
is chosen higher, provi-
sion has to be made to re-
duce the dosis.
For the radial reflector of
the gas turbine reactor,
the possibilities have
been discussed for a reduc-
tion of the dosis (/4/ pg.
12). In Fig. 27 the curve
3A gives the dosis tempera-
ture diagram for the design
without any provision. The
ideal way for a limitation
of the dosis is to replace the reflector at time intervals of 12 - 15 years
which yields curve 3. Another proposal is to insert a small amount of neu-
tron poison in the reflector graphite in the area of the highest exposure
(curve 2). This shifts the neutron flux profile towards the inner of the
core. A third alternative is to provide a third radial fuel element zone
of about 40 cm thickness, which is fuelled with balls of reduced fissile
65
and increased fertile material content (curve 3).
The top reflector is sufficiently shielded by the presence of the control
system in the cavity above the surface of the pebble-bed/ as discussed in
the following chapter.
4.12 Effect of the Upper Void
SIMULIERTER OBERER
II,! REFLEKTOR + EISEN
! REFLEKTOR' Mu
i!] HOHLRAUM
BEREICH '
nilMIIII
IIIIII
REGELRINGE
CORE
RE
FMHO
FIG. 28: Simulation of the area
above the core for trans-port calculation
Between the surface of the pebble-bed
and the top reflector there is a ca-
vity of approximately 1 m thickness
(Fig. 28). The straight flight of
the neutrons cannot be simulated by
diffusion theory but requires transport
or Monte Carlo method. It has been
derived by V. MALY /8/ that the local
flux distribution in the cavity is
almost constant. This results in a :
flat radial distribution at the upper
boundary of the pebble-bed.
FIG. 29; Thermal flux variation when inserting control poison into theupper cavity.
66
Insertion of control rods in the cavity effects in a global lowering of the
neutron flux level. This means that the presence of the control rods reduce
the flux preferably at the whole surface, and strong local changes of the
fuel element power output are excluded. A pessimistic 2-dimensional model
has been formed by MALY,which assumes the many control rods to be lumped to-
gether in 1 or 2 rings (Fig. 28). The grayness of these rings simulate the
insertion depth in the cavity. The three drawing of Pig. 29 illustrate the
corresponding effect.
By moving the rods above the surface of the pebble-bed, it is possible to
control about 2.6 nile (/4/ pg. 45). This is mainly caused by shifting of
the neutron flux profile towards the lower core with the high content of
accumulated fission products (/2/ pg. 60).
4.13 Reactor Control
At full power equilibrium cycle, a certain number of control rods must be
inserted in the upper cavity in order to allow reactor control for the curr-
ent load follow. Approximately 2 nile are required. In the used burnup mo-
del this is simulated by a corresponding homogeneous poisoning of the re-
flector area above the pebble-bed surface. Withdrawing of this poison brings
a change in the axial power density profile as given in Fig. 30 (/3/ pg. 49).
The corresponding change in the temperature distribution is also given in
this figure. The maximum of the fuel temperature is shifted towards the top.
It increases only by 38 °C. This low change is due to the lower temperature
of the surrounding gas and to the higher thermal conductivity as given in
Fig. 25. .
POWER AND TEMPERATURES AT THE AXIS WHEN REMOVING THE CONTROL POISON
5-7 -TECHN.LIMIT 0
100 200 300 400 500 CM
CENTER
FIG. 30:Change in power andtemperature distribu-tion at withdrawal ofcontrol poison.
67
In réf. /2/ (pg. 60) has been outlined that a long term presence of the
control poison in any special position influences significantly the equili-
brium distribution of the isotopes and of the power density in the core.
Nevertheless, the efficiency of changes in the amount of the control poison
(i.e. in the rod insertion depth in the void) stays almost unaffected.
4.14 Reactor Shut Down
TABLE XXIVShut-down reactivity changes
Fuel cycle
Average power density MW/m
Moderation ratio Nc/NHM
Gas temperature rise °C
Reference
keff:
Pull power —> 0-power, hot nile
Full power —* 0-power, cold nile
Long term decays, 0-power nile
Low enr.
5
350
250 —* 810
IV pg.42
2.2
8.5
0.4
Th/U-5
8
255
250 —* 770
IV pg.45
2.2
7.7
6.3
The reactivity requirement for shut-down depends strongly on the design data.
Up to now it has been calculated for two different cases which are listed
in Table XXIV. The long term requirement is considerably higher for the
thorium cycle because of the Pa-233 decay.
The primary shut down system provides absorber rods which are pressed or
screwed into the pebble-bed. The secondary one provides small boron contain-
ing spheres to be poured into the pebble-bed or, alternatively, a gaseous
absorber material.
H.J. NEEF /18/ has simulated the primary system by means of the 3-dimension-
al diffusion code CITATION. An application has been given by G. BALLENSIEFEN
/4/ (pg. 56) for the gas turbine reactor being closely related to case 4011
of the present report. Herefrom we have taken the Fig. 31 which illustrates
the influence on the neutron flux distribution in the core. Fig. 32 repre-
68
. i . i . i . 1 . i .
1 2 3 1 5
RADIUS (K)
FIG. 31: Thermal flux in r-& coordinates.A) Unrodded core; B) rods inser ted .
7 EINTAUCHTIEFE (H)
FIG. 32; Efficiency of shut down rods bank
69
sents the efficiency of the bank of absorber rods when.pressing into the
pebble-bed. The dotted curve shows that a penetration depth of 2.5 m brings
a2\ k „ = 9.8 nile, compared to the position of full power equilibrium
operation. The fat curve has been derived under the assumption of a poisoned
bottom reflector, which has been found by NEEF /18/ to be a useful aid for
improving the efficiency of the shut down rods. Here, a penetration depth
of 2.5 m might be sufficient for a short term cooling down or for a long
term shut-down under keeping the core hot, including an estimated engineer-
ing factor for the safety margin.
4.15 Xenon Distribution
XEU8N-OVERRI0E
1.0
o.n
POWER DIRGRAM
1.0
FIG. 33: Xenon Override
The reactor analysis programme V.S.O.P. (Ch. 1.5.2) calculates automatically
the xenon-override at given time steps of the reactor life. For a 100-40-
1OO % load follow, the build-up history of the Xe-135 is given in Fig. 33
for 11 equidistant points at the reactor axis (/2/ pg. 46). The well known
increase in xenon density takes place only in the upper regions. Here, the
rate of Xe-annihilation is preferably due to the amplitude of the thermal
flux, which is reduced proportionally to the power. In the lower core the
70
Xe-removal is mainly due to spontaneous decay. If decreasing the power, one
should expect an exponential decrease in the lower curves. But, instead, there
is a slight increase. This fact indicates that the build-up of the xenon
poison in the upper core shifts the flux and power to the lower core area.
In order to maintain the reactor critical, a certain response of the control
system is required which influences also the global power density distribu-
tion. This is not included here.
A more complete treatment can be performed by means of the code system
ASTERIX /23/ being specialized for this concern. On this basis a research
on possible xenon oscillations in large core units will be given in volume
11 of this report series. From the view of the present study three trends
can be expected:
1. Fig. 33 indicates an insensitivity of the xenon concentration in the
lower half of the core with respect to a neutron flux variation as
large as 60 %. This suggests to expect a damping influence on os-
cillations by the typical axial OTTO profile of the flux distribution.
2. This effect cannot be expected for a high conversion variant, because
the power is uniformly distributed over the core height.
3. The effect of the upper void is expected to restrict the local effi-
ciency of control rods, but on the other hand a damping of oscilla-
tions is expected from the flux equalization in the cavity, as des-
cribed in Ch. 4.12.
4.16 Temperature Coefficients
For all more detailed investigated concepts, temperature coefficients have
been calculated. The method consists in the comparison of the k of two
complete spectrum-diffusion runs, for which the temperature assumption diff-
ers by 100 °C all over the core. v-.
Doppler coefficients due to the resonance broadening of the fertile material
are found to vary between -4.0 and -4.5 # 10 ^ k/°C for the low enriched
cycle /2/, and between -2.4 and -3.3* 10 for the thorium cycle /3/, /4/.
The reactivity change due to the shifting of the thermal spectrum is gene-
rally smaller by one order of magnitude. For the thorium equilibrium core
71
/3/, /4/, small positive values have been observed, but the total tempera-
ture coefficient being predominantly due to the doppler effect is markedly
negative.
For one low enriched case /2/ pg. 45,the stepwise cooling down from 1000
to 300 °K has been evaluated in terms of reactivity and spectrum effects.
The U-238 resonance integral decreases by 19 %. The thermal averaged absor-
tion cross sections turn out to increase by 42 % for the U-235, to decrease
by 27 % for the Pu-239, and to swing by 10 % around the average value of the
full power operational value for the Pu-240 isotope.
Although the method of deriving the temperature coefficients is reliable,
this field requires more detailed research. The assumption of a parallel
temperature rise all over the core is very poor. For the reactor type under
consideration a strong local variation of the temperature changes caused by
the operational performance must be expected. It is recommended to study the
response of reactivity more in detail.
72
REFERENCES
/l/ HANSEN, ü., SCHULTEN, R. , TEUCHERT, E, :Physical Properties of the "Once Through Then Out" Pebble-Bed Reactor.Nuclear Science and Engineering: 47, 132 (1972).
/2/ .TEUCHERT, E., MALY, V., HAAS, K.A. :Basisstudie zum Kugelhaufenreaktor in OTTO-Beschickung.JUL-858-RG, Kernforschungsanlage Jülich GmbH (Mai 1972).
/3/ TEUCHERT, E., MALY, V., HAAS, K.A. :OTTO-Kugelhaufenreaktor im Thorium-Brennstoffzyklus.JÜL-1O59-RG, Kernforschungsanlage Jülich GmbH (April 1974).
/4/ TEUCHERT, E., BALLENSIEFEN, G. , HAAS, K.A., MALY, V., MIELKEN, G.,PETERSEN, K., RUTTEN, H.J., UHLENBUSCH, L., WILL, M., WOLF, L.:OTTO-Kugelhaufenreaktor für eine 1000 MWe Heliumturbinenanlage.JÜL-1070-RG, Kernforschungsanlage Jülich GmbH (Mai 1974).
/5/ TEUCHERT, E. , MALY, V., HAAS, K.A. :PR-3000-1, Gleichgewichtsbetriebszyklus.Internal Report IRE-34-73 (December, 1973).
/6/ TEUCHERT, E., MALY, V.:Comparison of the Thorium and Low Enriched Uranium Fuel Cycle in theOTTO Pebble-bed HTR. .DCPM 16/KFA 4, March 22, 1973. •
/!/ MALY, V., TEUCHERT, E.:Separated Location of the Partially Depleted Fuel in the Pebble-bedReactor.Reactor Burn-up Physics, Proceedings of a Panel, Vienna, July 12-16,1971, Pg. 57-70, Vienna (1973).
/8/ MALY, V.:Neutronenphysikalischer Einfluß des oberen Hohlraumes im Kugelhaufen-reaktor .JÜL-932-RG, Kernforschungsanlage Jülich GmbH (März 1973).Dissertation D 82 RWTH Aachen .
/9/ BOHN, T., KOMAREK, P., NAOMIDIS, À., NICKEL, H. , TEUCHERT, E. :Feasibility Problems on UHTR for Closed Cycle MHD Power Plants.Symposium on Engineering Aspects of Magneto-hydrodynamics, Stanford,March 26-28, 1973.
/10/ TEUCHERT, E., MALY, V.:Thermal and Neutronic Performance of Pebble-bed Fuel Elements.BNES Conference on nuclear fuel performance, London, October 15-19, 1973.British Nuclear Energy Society.
/ll/ MALY, V., SCHULTEN, R., TEUCHERT, E.:500 MWth Kugelhaufenreaktor für Prozeßwärme in Einwegbeschickung.Atomwirtschaft \1_, 216 (April 1972).
73
/12/ MALY, V., TEUCHERT, E.:
Einweg-Kugelhaufenreaktor zur Elektrizitätserzeugung.
Atomwirtschaft \T_, 518 (September 1972).
713/ TEUCHERT, E., WOLF, L. :
Das OTTO-Konzept für Hochtemperaturreaktoren.Energie und Technik 25, 236 (1973).
/14/ TEUCHERT, E.:Resonanzabsorption in einer zweifach heterogenen Anordnung kugelför-miger Brennelemente. INukleonik, H_, 68 (1968). '• |
/15/ HANSEN, U., TEUCHERT, E.:Influence of Coated-Particle Structure in Thermal-Neutron-SpectrumEnergy Range.Nuclear Science and Engineering 44_, 12 (1971).
/16/ MALY, V., SCHULTEN, R., TEUCHERT, E.:
Einweg-Kugelhaufenreaktor als Hochkonverter im Thoriumzyklus.Atomwirtschaft, 9 (12), 601 (1974). ' ; •
/17/ TEUCHERT, E., MALY, V. :Numerical Research on the Pebble-bed Reactor.ANS Topical Meeting, Atlanta, September 8-11, 1974.
/18/ NEEF, H.J.: f
Berechnung der Wirksamkeit von Absorberstäben in Hochtemperaturreak-toren unter Verwendung transporttheoretisch bestimmter Randbedingun-gen mit einem dreidimensionalen Diffusionsprogramm.JÜL-98O-RG, Kernforschungsanlage Jülich GmbH (Juni 1973).Dissertation D 82 RWTH Aachen.
/19/ HANSEN, U., SCHULTEN, R., TEUCHERT, E.:Some Physics Aspects of a Small "Once-Through" Pebble-bed Reactor.Trans. Am. Nucl. Soc. , 13(2), 834 (1970).
/20/ HANSEN, U., TEUCHERT, E.:V.S.O.P. Reactor Life Code System.To be published as JÜL-Report in near future.
/21/ BINKELE, L.:Ein Verfahren zur Bestimmung der Wärmeleitfähigkeit von neutronenbe-strahlten Graphiten bei Temperaturen zwischen 50 und 1000 °C.JÜL-1O96-RW, Kernforschungsanlage Jülich GmbH (August 1974).Dissertation D 82 RWTH Aachen.
/22/ ROTTEN, H.J.:Numerische Untersuchungen zu einem 2700 MWth OTTO Kugelhaufenreaktorunter den Betriebsbedingungen eines Heliumturbinenkraftwerkes.JÜL-1141-RG, Kernforschungsanlage Jülich GmbH (Dezember 1974).Dissertation D 82 RWTH Aachen.
/23/ LAUER, A.:Räumliche Xenon-Schwingungen in Hochtemperaturreaktoren.JÜL-85O-RG, Kernforschungsanlage Jülich GmbH (Mai 1972).Dissertation D 82 RWTH Aachen.