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SAFETY SER IES

No. 28

Management of Radioactive Wastes

at Nuclear Power Plants

INTERNATIONAL ATOMIC ENERGY AGENCY

VIENNA, 1968

This publication is no longer valid Please see http://www.ns-iaea.org/standards/

MANAGEMENT OF RADIOACTIVE WASTES AT NUCLEAR POWER PLANTS


The following States are Members o f the International Atom ic Energy Agency:

AFGHANISTAN GERMANY, FEDERAL NORWAYALBANIA REPUBLIC OF PAKISTANALGERIA GHANA PANAMAARGENTINA GREECE PARAGUAYAUSTRALIA GUATEMALA PERUAUSTRIA HAITI PHILIPPINESBELGIUM HOLY SEE POLANDBOLIVIA HUNGARY PORTUGALBRAZIL ICELAND ROMANIABULGARIA INDIA SAUDI ARABIABURMA INDONESIA SENEGALBYELORUSSIAN SOVIET IRAN SIERRA LEONE

SOCIALIST REPUBLIC IRAQ SINGAPORECAMBODIA ISRAEL SOUTH AFRICACAMEROON ITALY SPAINCANADA IVORY COAST SUDANCEYLON JAMAICA SWEDENCHILE JAPAN SWITZERLANDCHINA JORDAN SYRIAN ARAB REPUBLICCOLOMBIA KENYA THAILANDCONGO, DEMOCRATIC KOREA, REPUBLIC OF TUNISIA

REPUBLIC OF KUWAIT TURKEYCOSTA RICA LEBANON UGANDACUBA LIBERIA UKRAINIAN SOVIET SOCIALISTCYPRUS LIBYA REPUBLICCZECHOSLOVAK SOCIALIST LUXEMBOURG UNION OF SOVIET SOCIALIST

REPUBLIC MADAGASCAR REPUBLICSDENMARK MALI UNITED ARAB REPUBLICDOMINICAN REPUBLIC MEXICO UNITED KINGDOM OF GREATECUADOR MONACO BRITAIN AND NORTHERNEL SALVADOR MOROCCO IRELANDETHIOPIA NETHERLANDS UNITED STATES OF AMERICAFINLAND NEW ZEALAND URUGUAYFRANCE NICARAGUA VENEZUELAGABON NIGERIA VIET-NAM

YUGOSLAVIA

The Agency's Statute was approved on 23 October 1956 by the Conference on the Statute of the IAEA held at United Nations Headquarters, New York; it entered into force on 29 July 1957. The Headquarters of the Agency are situated in Vienna. Its principal objective is "to accelerate and enlarge the contribution of atomic energy to peace, health and prosperity throughout the world".

© IAEA, 1968

Permission to reproduce or translate the information contained in this publication may be obtained by writing to the International Atomic Energy Agency, Karntner Ring 11, A-1010 Vienna I, Austria.

Printed by the IAEA in Austria December 1968


SAFETY SERIES No. 28


INTERNATIONAL ATOMIC ENERGY AGENCY VIENNA, 1968


MANAGEMENT OF RADIOACTIVE WASTES A T NUCLEAR POWER PLANTS (Safety Series, N o .28)

ABSTRACT. A manual prepared on behalf o f the A gency at a m eeting held in Vienna, from 29 April to 4 May 1968, by four consultants from four o f the M em ber States having the greatest experience in the generation o f e le c tr ica l power from nuclear stations. T w o representatives o f the World H ealthO rganizationr who also attended the m eeting , contributed a chapter on Standards and criteria .

Contents: I. Summary o f designs and operating experiences; II. Factors to be considered in design and operation; III. Future trends and perspectives; IV. Tables I-1V ; References; Appendixes: 1. Radioactive waste m anagem ent at Canadian nuclear power reactors; 2 . Radioactive waste m anagem ent at nuclear power plants in France; 3 . The m anagem ent o f radioactive wastes from com m ercia l nuclear power stations in the United Kingdom ; 4 . United States practice in m anagem ent o f radioactive wastes at nuclear power plants.

Entirely in English.

(225 p p . , 14.8 x 21 cm , paper-bound, 20 figures; 1968) Price: US$6.00; £2.10.0


IAEA, VIENNA, 1968 . STI/PUB/208


FOREWORD

The International Atomic Energy Agency has produced this manual in response to a number of requests for information prompted by the rapid expansion of the nuclear power industry. It is limited to a discussion of radioactive wastes that arise at nuclear power stations, and does not deal with wastes that result from the reprocessing of nuclear fuel. The manual was prepared at a meeting held in Vienna from 29 April to 3 May 1968 by four consultants from four of the Member States having the greatest experience in the generation of electrical power from nuclear stations. Two representatives of the World Health Organization contributed Section II.2. on Standards and Criteria and the representative from EURATOM translated into English Section II. 4 on Process Capabilities, written by the French consultant.

The consultants for the project were:

Mr. Morton I. Goldman Vice President and General Manager,(Chairman) Environmental Safeguards Division,

NUS Corporation, Washington, D.C., United States of America

Mr. C.A. Mawson Head, Environmental Research Branch, Biology and Health Physics Division, Atomic Energy of Canada, Limited, Chalk River, Ontario, Canada

Mr. G. Cohendy Chef de la Section des Operations Radioactive, Centre de Marcoule,Bagnols s/Ceze (Gard), France

Mr. I. Dougall Nuclear Health and Safety Department, Central Electricity Generating Board, London, United Kingdom


The representatives of international organizations were:

Mr. R. Leupold Mr. H. Schmier

World Health Organization, Geneva, Switzerland

Mr. J. Van Caeneghem Direction Generale Industrie et Economie, EURATOM, Brussels, Belgium

In addition to the body of this manual, the consultants prepared review papers on the management of radioactive wastes at several nuclear power stations in their countries; these have been included as Appendixes. Much of the detailed description and data in these review papers were contributed by personnel at the reactor sites.

The project was organized and directed by Mr. FrankN. Browder of the IAEA as Project Officer and Scientific Secretary of the consultants' committee.


CONTENTS

I. SUMMARY OF DESIGNS AND OPERATING EXPERIENCES 1

II. FACTORS TO BE CONSIDERED IN DESIGN ANDOPERATION................................................................................. 5

II.1 .Sources and character of wastes ............................................ 7II. 1.1. Basis of determination ................................................ 7II. 1.2. Potential sources of radioactive wastes .................. 9

II.2 .Standards and criteria .............................................................. 13II. 2. 1. International recommendations.................................. 13II. 2. 2. National standards........................................................ 15II. 2. 3. Design guidance .................................................. . 16

II .3. Environmental factors .............................................................. 17II. 3. 1. In-Plant .......................................................................... 17II. 3. 2. Off-Site .......................................................................... 19

11.4. Process capabilities.................................................................. 23II. 4.1. Relationship to management philosophies................ 23II. 4. 2. Relationship to process costs .................................... 28

11.5. Performance verification and monitoring ............................ 29II. 5.1. Role of in-plant controls.............................................. 30II. 5. 2. Role of environmental monitoring ............................ 32

III. FUTURE TRENDS AND PERSPECTIVES.............................. 35

IV. TABLES I-IV .............................................................................. 43

REFERENCES....................................................................................... 56

APPENDIXES

Appendix 1. Radioactive waste management at Canadiannuclear power reactors ..............................C . A . Ma w s o n

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Appendix 2. Radioactive waste management at nuclearpower plants in France ............................................ 77G . C o h e n d y

Appendix 3. The management of radioactive wastes from commercial nuclear power stations in theUnited Kingdom .......................................................... 93I . D o u g a l l

Appendix 4. United States practice in management ofradioactive wastes at nuclear power plants.........147M . I . G o l d m a n


SUMMARY OF DESIGNS AND OPERATING EXPERIENCES

(Section I)


I. SUMMARY OF DESIGNS AND OPERATING EXPERIENCES

Design data and operating experience with waste-management systems at nuclear power stations in Canada, France, the United Kingdom and the United States of America are described. Although the specific designs and operating practices vary between nations, the underlying philosophies are essentially identical, being based on the protection principles of the International Commission on Radiological Protection (ICRP). Similarly, the operating experience with these different waste-management systems has been uniformly excellent in maintaining the radiation exposures in the environment well below the accepted dose standards.

Design and operating experience is provided in the Appendixes by each of the participating countries, and reference may be made to these for specific details of interest. In this section and those which follow, the general design practices and operating principles common to all facilities are presented in summary fashion (Tables I-IV), as well as a discussion of appropriate standards and criteria, environmental factors to be considered, process capabilities and performance verification and monitoring. An examination of future trends and perspectives is also included.

Tables I - IV will be found in Section IV before the Appendixes.

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FACTORS TO BE CONSIDERED IN DESIGN AND

OPERATION (Section II)


II. FACTORS TO BE CONSIDERED IN DESIGN AND OPERATION

II. 1. Sources and character of wastes

II.1.1. Basis of determination

The identification of the sources, characteristics and quantities of the wastes is a basic preliminary to the design of the waste-management system. The capacity and flexibility of the system ultimately provided will depend on various factors both technical and economic, but in the first instance the designer will need to make a realistic appraisal of the likely occurrences. The information re quired may be stated as:

Types of wasteQuantities of wastePhysical and chemical characteristicsRadioactivity levels.

II. 1.1.1. Types and quantities. The main types of waste and the quantities arising in normal operation can generally be estimated from consideration of the reactor and ancillary plant design; preferably, reference should also be made to operating experience at power stations of similar type (see Section IV, Tables I-IV), a lthough the revelance of such experience depends on the intended mode of operation of the new station compared with existing stations.

Account should be taken of possible changes in the usage of the station during its lifetime, e.g. from "base load" to "load-following" use. In general, variations in plant power levels will produce larger volumes of wastes than operation at a steady power level.

Throughput capacity of waste systems will be based on "maximum" occurrences in normal operation, and therefore "batch" quantities should be estimated as well as average annual occurrences. In addition, the system capacity must include allowance for occasional abnormal occurrences of waste from minor incidents (operational mischances), which invariably occur from time to time.

Experience at many power stations has been that the quantities of solid wastes of very low radioactivity content produced have considerably exceeded design estimates; reference to the summarized data (Section IV, Tables I-IV) on operational experience is therefore recommended when considering the requirements for such wastes.

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11.1.1.2. Physical and chemical characteristics. Knowledge of the physical and chemical nature of the wastes is of considerable im portance; for solid wastes, characteristics such as pyrophoricity, dust loading and chemical reactivity will influence the procedures for collection, treatment, accumulation and/or disposal. As examples, highly dust-loaded solid wastes may require special precaution in collection, treatment (e.g. by baling) and accumulation. The presence of non-combustible components may preclude volume reduction by incineration. Likewise the production of acid fumes by incineration of materials such as PVC will affect incinerator design.

Some fuel-element cladding materials are pyrophoric in finely divided form, and safety precautions are required in transporting or accumulating them.

Chemical incompatibility of solid wastes may lead to requirements for segregation.

In the case of liquid wastes, acid or alkaline liquors may re quire segregation from other wastes for neutralization before treatment or disposal. Similarly, it will sometimes be desirable to collect high-purity liquid wastes separately, to allow their recovery and re-use after treatment. Decisions on filter types and provision for ion exchange, flocculation and evaporation will also depend on the chemical conditions of the wastes as well as on their radioactivity levels and volumes. Characteristics of gaseous wastes such as pressure, temperature, oil or moisture levels, and explosive or corrosive gas content may impose special requirements on treatment and disposal systems.

11.1.1.3. Levels and nature of radioactivity. The importance of accurate knowledge of the levels and nature of radionuclides in waste depends on the volume of the waste expected and the order of activity levels in the waste. Generally, where disposal to the outside environment is proposed, such information is essential.

The provision of the information will require an assessment of the levels of neutron activation products in the various materials (e.g. fuel cladding, coolant, moderator, etc.) within the reactor and primary shield. In addition, the various mechanisms by which these become transferred from the reactor, i.e . in clean-up and fuel-handling systems, will have to be assessed. The assessment of radioactivity levels should also take account of the possible presence of fission products released from defective fuel. This will

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require some standard assumptions to be made regarding the type and frequency of fuel failures in the reactor, fuel-handling and decay- storage systems. These assumptions should, again, where possible, be based on operational data and should take account of the possible effects of changes in fuel irradiations from original design targets.

II. 1.1.4. Applications of estimates. These include:

(i) Design of plant for waste management, manpower requirements;

(ii) Design of monitoring equipment (e.g. indication of isotopes for which special detection equipment will be needed);

(iii) Estimation of capacity of external environment for r e ceiving wastes;

(iv) Design of environmental monitoring program s;(v) Safety submissions and applications for licence to operate.

II. 1.2. Potential sources of radioactive wastes

II.1.2.1. Main types of waste from heavy-water reactors (Canada)

(a) Solid wastes (in three categories):

Category 1 < 200 mR/h on contact

Contaminated scrap metal and plastic,Rags, paper and wood,Used air filters,Sump dregs.

Category 2 200 mR/h to 1 R/h on contact

Scrap and crud from fuelling machines and process systems.

Category 3 > 1 R/h on contact

Ion-exchange resins,Reactor components (from replacement of pressure tubes). Spent fuel assemblies.

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(b) Liquid wastes:

Organic liquids from control laboratories and decontamination work,Aqueous liquids from laundry and washrooms,Slurry from storage bay for spent fuel,Low-concentration heavy water.Aqueous wastes from control laboratories, decontamination work, building floor drains and subsurface drainage.

(c) Gaseous wastes:

Thermal-shield-cooling system bleed,Excess volume and purges from closed systems (vaults), Off-gases from .Primary-Heat-Transport System,Air from accessible work areas in reactor building,Flow from fume hoods in laboratories and decontamination centres,Air from waste-management areas and spent-fuel storage rooms,Air from active workshops.

II.1.2.2. Main types of waste from graphite-moderated, gas-cooled reactors (United Kingdom)

(a) Solid wastes:

Debris from fuel element or stringer dismantling,Liquid filter sludges,Spent-liquid filter materials (sands, precoat). Spent-ion-exchange materials,Ventilation filters,Protective clothing,Protective floor coverings (temporary),Paper towels, hats, etc.,Swabs,Reactor-coolant-circuit spent-filter elements, Reactor-coolant-circuit spent-filter dusts, Spent-shield-cooling air filters (steel pressure-vessel re actors only),Dessicant from reactor-coolant drying plant.

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Incinerator ash,Control rods,Flux flattening bars,Charge chutes.Shield plugs,Reactor instruments, e.g. ion chambers,Fuelling-machine grabs.

(b) Liquid wastes:

Personnel changing rooms - showers, washbasins,Laundry wastes,Fuelling-machine washdown,Plant decontamination,Floor decontamination,Ion-exchange resin regenerants,Filter backwash,Spent-resin and filter-transfer liquors,

* Pond purge,Reactor-coolant drier-plant liquors.

(c) Gaseous wastes:

Shield-cooling air (steel pressure vessel reactors only), Ventilation air,Reactor-coolant (C02) venting (e.g. from main pressure- circuit-fuelling machinery, BCD systems and gas circulator seals),Reactor-coolant leakage,Pressure-vessel-cooling-system venting (concrete pressure- vessel reactors only),Incinerator off-gasies.

II.1.2.3. Main types of waste from natural uranium graphite gas reactors (France)

(a) Solid wastes:

Wastes from decontamination (cotton, vermiculite, etc.), Safety gloves (linen and rubber),Air filters,

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Small contaminated tools,Irradiated pieces of metal,Scraps or strips of PVC or polyethylene,Paper of various kinds and other wastes.

(b) Liquid wastes:

Sludge effluent,Water from scrubbers,Equipment decontamination water,Water from showers;Oils ,Solvents.

(c) Gaseous wastes:

Ventilation air,Coolant (in the event of core drainage with discharge into the atmosphere).

II.1.2.4. Main types of waste from light-water reactors (United States of America)

(a) Solid wastes:

Spent resins.Air filters & liquid filters (cartridge or precoat), Evaporator bottoms.Laboratory glassware.Protective clothing (gloves, overalls, etc.),Tools and equipment,Miscellaneous paper and other wastes.

(b) Liquid wastes:

Reactor coolant (i) Expansion overflow (start-up),

(ii) Leaks (valve stems, pump seals, etc.),(iii) Samples (for laboratory analysis),(iv) Letdown (for chemical reactivity adjustment)Floor drains.

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Laundry wastes.Decontamination wastes (tools, glassware, etc.),Resin regenerant solutions,Resin transfer water,Personnel decontamination (showers, etc.).

(c) Gaseous wastes:

Reactor coolant vents (pressurized-water reactor),Main condenser air ejector (boiling-water reactor), Liquid-waste tank and equipment vents.Ventilation air.

II.2. Standards and criteria/

11.2.1. International recommendations

11.2.1.1. Primary and derived standards. On an international basis, the primary guidance for radiation dose limits (for workers, individual members of the public and the whole population) has been given by the International Commission on Radiological Protection (ICRP). The ICRP recommendations are periodically reviewed as more information becomes available, the most recent recommendation being contained in their Publication 9 [1]. Using the primary standards (or dose lim its), it is possible to calculate for a given radionuclide the intake for standard man (ICRP Publications 2 and 6) [2,3], which would during its passage through and partial retention in his body or a portion thereof deliver the dose limit. The portion of his body receiving the relevant limiting dose is called the "critical organ" and the calculated intake represents a secondary standard for controlling his exposure. Using this approach, the IAEA has calculated the values of maximum permissible annual intakes for members of the public, both for ingestion and inhalation, for a large number of radionuclides. These are published in its 1967 Revised Basic Safety Standards (Safety Series No. 9) [4]. Additional guidance, which is for the most part derived from the ICRP recommendations, may be found in reports such as the World Health Organization International Standards for Drinking Water [5].

As explained above, the secondary guides, so calculated, are based on a "standard man", which is an artificial model obtained by averaging certain anatomical and physiological characteristics

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of an "average" adult person. In general, the quoted concentrations apply only to individual radionuclides in either air or water, and are for continuous inhalation or ingestion by the average adult. Directly applicable recommendations cannot always be found from the international sources listed, especially when the "critica l groups" are children. In addition, the guides do not take into consideration the extremes in intake above and below the average or the chemical form of the radioactive materials (except in respect to their solubility). ICRP Publication 7 [6], however, gives details of the significance of variations within a critical group.

II.2.1.2. Application to waste management. From the point of view of the control of the general environment, radioactive, materials (critical nuclide(s)) can enter the body (critical organ) from any one of or a combination-of the following environmental media (critical pathway): air, water and food. In any specific situation, before an appropriate release guide can be developed, the relative contribution to total intake from each source must be investigated and taken into account. As more radionuclides are observed in the mixture to be released, and particularly if they cause exposure of different body organs, direct application of the recommended maximum permissible concentration becomes more difficult and complex since it is at present uncertain to what extent the effects of irradiating several organs simultaneously are simply additive. As a general guide, ICRP Publication 9, paragraph 68, states that " if three or more body organs are each receiving more than one- half of their respective maximum permissible doses, the exposure shall be regarded as excessive".

Added to these complications is the fact that experience has shown that permissible levels for discharge do not depend solely upon recommended intake values. Perm issible levels for each segment of the environment must be derived on the basis of studies of local conditions. For example, concentrations of radionuclides in air and water, which are too low to be of concern from the point of view of direct inhalation or ingestion, may lead to appreciable contamination of foods. Thus, in the case of release to the aquatic environment, downstream uses may place more stringent requirements on perm issible concentration than the limitations imposed by use of the water for drinking purposes. Also, if radioactive materials are already present in the environment into which d isposal is planned, the permissible concentrations or calculated rates

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of release will have to be modified to take into account the overall exposure situations.

Although individual evaluations will have to be made for essentially all nuclides being released to the environment, many of them can be rapidly eliminated as a significant contributor to (population) dose simply on the basis of theoretical assessment. In fact, such an approach will generally narrow the number down to only one, or at the most a few, radionuclides for which sampling, analysis and study need be performed (critical nuclides).

In conclusion and as previously mentioned, before releasing, radioactive material it is necessary to determine which nuclides passing through which pathways lead to exposure of which "critical" population group.

II.2.2. National standards

The recommendations of the ICRP have no direct legal connotation in themselves but do provide and serve as the generalbasis for the development of most national regulations and standards. In fact, many countries in the past have found this to be useful. That is, they set the limiting concentration in the receiving air or water at one tenth (or some other fraction) of the maximum permissible concentration for continuous intake for the radionuclide under conditions of occupational exposure. One of the major faults in using this approach is that it does not allow for subsequent concentration of radionuclides in food materials later to be ingested by the population at risk. For example, it is well known that over pasture land, permissible discharges of 131I to the atmosphere should be established not on the basis of inhalation but mainly for milk from cows grazing the pasture.

While mentioning this exception, it should also be pointed out that in many other cases application of the ICRP maximum per - missible concentrations has led to acceptable conditions for effluent release control. If initial control levels are thought to be too re strictive from the point of view of the economics of the nuclear plant, there will be an incentive to conduct "critical-path" studies that may demonstrate the feasibility of raising discharge levels safely.

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Some countries use arbitrary concentrations as a basis for regulatory control of effluents but in any case it should be sufficient that those responsible for the health and safety of the public be aware of the problem and look towards an eventual change in the course of time when sufficient information has been gathered. Examples of different types of approaches to this problem are given in the IAEA publication, "Basic Factors for the Treatment and Disposal of Radioactive Wastes"[7].

II.2.3. Design guidance

The systems installed at nuclear power stations to collect, process, handle, store and dispose of radioactive wastes must be capable of meeting the requirements of the international and/or national standards and criteria discussed above. However, these standards almost uniformly recommend or require the maintenance of radiation exposures at the minimum practicable values. The question is often raised by engineers as to what design objective should be used to evaluate process or system performance requirements. This question is unfortunately not answerable in a straightforward manner.

The establishment of such a design objective for a particular station must take into consideration a number of factors including: the degree of realism (or uncertainty) in the estimates of waste generation, both in volumes and activities; the confidence of the engineer in the performance of the waste-processing system; the uncertainties in environmental transport and dispersion pathways to man; the degree of assurance as to future uses of the plant environment; and the interactions with other sources of exposure, either existing or possible in the future, i.e. additional nuclear generating units at the same site, nearby stations or stations on the same body of water, and other nuclear facilities such as spent-fuel reprocessing or re search installations.

If these factors (and perhaps others) can be identified with assurance in a quantitative fashion and included in the consideration of design, no further margin may be required. However, as in any engineered system, it is always prudent to reserve some margin of safety to cover unforeseen problems, the magnitude of this margin depending upon the informed judgement of the parties concerned.

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II.3. Environmental factors

11.3.1. In-plant

11.3.1.1. Handling methods. Methods of handling wastes must be designed to keep the radiation exposure of workers to a practical minimum. This implies the use of distance, shielding and time in the operating procedures and in the design of the plant.

Distance: In the design process the available space in a power station is mainly devoted to the reactor and its components, and to the turbine and its auxiliary equipment. The waste-management department tends to have a low priority in the mind of the designer, which may result in the crowding of the facilities into inadequate space.

It must be realized that radiation fields in this department will be variable and may be quite high, especially after an incident (spill or operational malfunction). Sufficient space must be left round tanks, pumps, filters and other fixed equipment for normal maintenance and movement of haulage vehicles. It must also be borne in mind that remote-controlled manipulators may be required in some places, and space must be left for their operation.

The waste-management staff is sometimes faced with the problem of temporary storage of waste, and a place should be reserved for this purpose remote from regular working areas and suitable for the erection of temporary shielding. In the general plant area there are some places where a temporary accumulation of decontamination materials or radioactive objects may be foreseen, and space should be available there.

Shielding: During the dismantling or maintenance of radio - active equipment or the cleaning-up of spills it may be necessary to work from behind shielding. The provision of portable shielding is necessary, with means for moving it, both in the waste-management area and in the rest of the plant. Parts of the structure known to be radioactive, such as pipes and tanks, should be located so that normal working areas are shielded.

Time: In designing the layout of the system, attention must be given to ensuring that the staff spends as little time as possible in "active" areas. The location of a sampling point at one end of a tank or at the other may make a difference to exposure of staff if they must pass beside the radioactive tank to take samples. Poor location of filters may unduly prolong the time necessary to change

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them. A time-and-motion study of the activities of the staff may enable the designer to reduce radiation exposures significantly.

Handling techniques: The worker must be protected from direct radiation and also from surface contamination and breathing hazards. Protection from direct radiation is obtained by use of remote handling equipment, shielding, skilful layout of plant, and careful area and personnel monitoring. Posting of working-time limits by competent radiation surveyors can do as much as anything else to r e duce direct radiation hazards.

Surface contamination is dangerous to the worker, to his colleagues and sometimes to his family as well as to the instrumentation of the plant. It can be avoided by using suitable protective clothing, efficient change-room operation, good personnel monitoring, and above all by well-planned handling techniques. Wastes should be placed in containers and sealed at the point of origin. Large objects should be wrapped in polythene and sealed. Trolleys and fork-lifts should be decontaminated after use, and traffic through the area should be so arranged that contaminated objects are not stored in working areas. When it is necessary to store wastes for significant periods before collection or to allow decay of short - lived contamination, a special storage place should be made available, designed for convenient access but shielded with respect to working areas.

In a large plant where there are several normal collection places for solid wastes or liquids in containers, a regular routine for collection should be set up so that people know when and where they can get rid of such wastes. It is often advantageous to have two collection bays at each pick-up point, one of them painted distinctively and labelled "ACTIVE". All wastes having a radiation field above a certain limit, e.g. 100 mR/h at 25 cm, are put in the "ACTIVE" bay. This helps the collection crew to avoid overexposure and to put the wastes in the right place for disposal.

After use, the disposal truck should be monitored and gross contamination removed. This will avoid tracking of spilled contamination all over the plant.

Protective clothing, ranging from rubber gloves, a laboratory coat and overshoes to a complete protective suit with an independent air-siipply, must be available and appropriate to the work being done. If tritium is present, it must be remembered that skin absorption is as important as the breathing hazard. Gas masks should

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be available to personnel, each of whom (because of the difficulty of fitting masks) should have his own.

11.3.1.2. Environmental controls. Internal contamination can be greatly reduced by a good ventilation system. It must be recognized that the system may not work as the designers intended, especially if doors are casually opened and closed. The direction of air movement should be checked under varying conditions by smoke tests or alternative methods (velometers) both in the plant and in the department, and where nearby doors open to the outside, with different wind directions. In places where storage of wastes may give rise to an airborne hazard, ventilation ducts should be available. Where a temporary airborne hazard exists (such as a spill), a plastic tent connected by hose to the ventilation system is useful.

Fixed monitoring instruments which alarm at a set level and provide a continuous record are useful for detecting general radiation and air contamination, but in addition intelligent monitoring by the health physics staff or the operators is needed. Personal monitoring on leaving the department should be obligatory, and bio- assay may be needed from time to time.

A changing room with adequate showers should be provided. Temporary changing rooms are also frequently useful in areas where local and transient contamination occurs.

11.3.2. Off-site

II.3.2.1. Land Usage. Land which is very valuable for agriculture or urban use cannot be used economically for disposal of wastes, and land that has received wastes can seldom be used for any other purpose unless the activity is very low or decays rapidly. The burial of low-activity wastes in open trenches uses up a large amount of land but engineered facilities for higher-activity wastes (such as concrete trenches, lined holes and "monoliths") occupy a relatively small area.

Waste-management problems may differ according to whether the reactor is in a rural or urban area. For example, in the case of gaseous wastes, the critical exposure routes at rural sites may be with agricultural products, whereas in urban areas inhalation or external radiation considerations will usually be more important.

For waste management, these considerations only apply to gaseous wastes, and for nuclear power stations the normal rate of

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release of radionuclides is quite low. For the establishment of waste- management areas for burial or other permanent disposal of solid wastes, land not otherwise valuable to man is likely to be used, but this may be contiguous to agricultural, urban or industrial areas.

The basic concepts and techniques used in ground disposal of wastes are dealt with in the IAEA Safety Series No.15 [8]. It need only be said here that in the less adsorptive soils the rate of movement of 90Sr through the soil is commonly 25 times slower than the rate of movement of ground water, and a figure of 100-1000 is more usual. Strontium is often the most rapidly moving of the more hazardous radionuclides. Local conditions will determine whether land burial is a suitable procedure.

II.3.2.2. Water usage.

(i) Direct consumption: Reactors located near bodies of freshwater are likely to discharge liquid wastes into water that will beused later for drinking. Whether the water is in a river or a lake,it will have a certain net transport rate past the point of discharge.This, together with the rate of mixing and the total volume of uncontaminated water available for dilution, determines the amountof radioactive material that can be put safely into the water in agiven period. Depending upon the body of water and the location ofintake and discharge structures, recirculation through the plant ofdischarged radioactive materials may also require consideration.

If there is a drinking water intake near to the effluent pipe fromthe plant, the concentration of radioactive material in the pipe maybe important. Beyond the point where significant mixing into themain body of fresh water ceases, only the amount released per unittime is important. The dilution capacity of the river or lake isusually enormous compared with the dilution that could be achievedby the plant operator by adding clean water to his effluent. So faras the river or lake is concerned, removing water and mixing inthe plant achieves nothing beyond the point where the effluent ismixed by nature into the main body of water. For this reason, thedesigner must think in terms of curies per day, not in microcuriesper millilitre.

Before deciding upon the permissible discharge into a body ofwater, it is important to find out what the water is used for. Thereare well-established standards for permissible concentration indrinking water, but if the water is used for fishing, swimming,

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industrial processing or irrigation, the drinking water standards are not applicable and the particular situation must be studied to arrive at the correct permissible discharge rate.

(ii) Indirect consumption: When the sea or other non-potable water is used for disposal or if the preliminary environmental investigation has shown that the "critical group" is not the people who drink the water, a study must be made which is best illustrated by an example. Let us assume that the discharge is into a river that supports a commercial fishery. It will probably be found that the "critical group" is the fishermen and their families, because the fish caught in any one place are usually "diluted" in the market by fish from other places.

Investigation shows that the average fish flesh consumption by members of the critical group is a g /d . The maximum permissible intake of a "critica l radionuclide" is b /uCi/d. The concentration factor between water and fish flesh is JT1. Thus the permissible concentration in the water is b /aF pCi/m l. The volume of water into which one day's effluent will be diluted at the site of the fishery is V ml. Hence, the permissible discharge per day is Vb/aF /^Ci. It must be emphasized, however, that the ICRP recommend control of dose to the public on an annual basis.

There are some circumstances in which radioactive contamination of mud or silt is significant, e.g. contamination of fishing gear or bathing beaches.

A practical evaluation of typical waste disposal problems and recommendations concerning the disposal of solid radioactive wastes into the sea is presented in IAEA Safety Series No.5, Chapter VI [9].

II.3.2.3. Dilution capability. In many cases the volume V can be assumed to be the total flow of a river per day, but in the case of a lake this parameter will not be so easily determined. Methods are available for obtaining a reasonably accurate figure on an annual basis, which is all that is necessary for satisfaction of the ICRP recommendations, but the expense of the necessary hydrographic work may suggest the use instead of very conservative assumptions.

If the water is used for any purpose by man between the effluent discharge point and the point of complete or sufficient mixing, the

i _ M C i/g flesh” ( jC i /m l w ater s im p le eq u ation s g iv en h ere n o a llo w a n ce is m a d e for

ra d io a c t iv e d e c a y . )

21


dilution regime should be determined experimentally by methods such as fluorimetric estimation of Rhodamine B dye dilution. It is important that the dilution pattern should be observed at the site, especially in a lake, because we are only concerned with local effects, and broad generalizations such as the directions of currents marked on hydrographic maps are almost useless.

Experience shows that in large rivers and especially in lakes, the inshore current depends greatly on the wind, which can move large masses of water in unexpected directions. Thus the prevailing wind and the configuration of the shore can produce a local dilution regime which would not be expected from examination of a large- scale map.

The designer will seldom have all the desirable information available when decisions have to be made — for example his d ischarge will be warm and the discharge itself may be sufficient to affect local currents. Warm water can form a surface layer which is poorly mixed with the remainder of the water body. However, competent hydrographic analysis can make allowance for these effects if the fundamental regime of the body of water is known.

Similarly, determination of the micrometeorological characteristics of the site are important in determining the acceptable r e lease rate of radionuclides by way of plant vents or stacks. The data to be collected will vary with the type of station and the relative importance of gaseous wastes, but will in most cases include measurement of the frequency distributions of wind direction, wind speed and atmospheric stability (either by lapse rate or by wind speed or direction variance). It is important that the measurements reflect the behaviour of the atmosphere over both the vertical extent and the distances of significance, particularly if topographic features are pronounced.

For some sites it may be sufficiently important that smoke or other tracer tests be conducted to confirm trajectories and dispersion conditions. Precipitation patterns may also be significant if materials subject to "washout" are to be released to the atmosphere from the station. More information on dispersion parameters is given in Techniques for Controlling Air Pollution from the Operation of Nuclear Facilities [10].

When the hydrographic, meteorological and demographic investigations have been completed, it will often be found that the maximum permissible discharge is large compared with the discharge rate normally expected from a nuclear power station. It is at this

22


point that the designer must recall the ICRP injunction to restrict radiation to man to the lowest practicable level, and also to r e member that his station will probably not be the only one to be erected near to that body of water or in that airshed. The permissible discharge that he has calculated is a maximum. The calculations, however, can only be expected to give an approximate answer and safety factors are usually introduced for this reason. The licensing authority may well restrict the discharge to a fraction of the theoretically permissible amount, but the use of unreasonable safety factors in setting limits of discharge should be avoided. This does not, however, relieve the operators of the duty to ensure that the radiation dose to the population should be kept to a minimum.

Dispersion of effluent gas from stacks located in most places other than deep valleys will give a received concentration (Ci/m3), averaged over a year, at distances of 1km or more, which is very unlikely to exceed 10 '6 s/m 3 multiplied by the emission rate from the stack (C i/s). This experimental observation is at least as r e liable for the purpose of setting limits as the use of the conventional micrometeorological equations.The equations cannot be used legitimately without considerable investigation to determine the parameters applicable at the locality, and are of doubtful value over an extended period [11].

II.4. Process capabilities

Each reactor type produces its particular types of radioactive wastes, and the processes used for waste treatment are related to the type of radioactive waste produced and to the particular sitesituation as far as the release of the radioactive affluents is concerned. In addition to the generation of radioactive wastes, the designer and operator of a nuclear power station must also consider industrial and sanitary wastes and heat rejected from the station. However, these are not within the scope of this manual.

II.4.1. Relationship to management philosophies

The basic principle of waste management is that the workers and the public should not be exposed to radiation levels in excess of the norms recommended by the ICRP. In practice, an additional safety factor is applied to these norms, so that maximum exposures are well below the recommended figures.

23


The established exposure limits can be met in various ways based on two different objectives:

dilution (dispersion) in the environment, orconfinement (with or without prior concentration).

As a function of their physical nature, three types of radio - active wastes are to be considered: solid, liquid and gaseous. For each of these categories the management consists of three major steps: collection, treatment and storage (or release).

II.4.1.1. Solid wastes. These wastes are produced in many different forms that lead to various treatment or storage methods. The final destination given to a particular waste is a function of its activity, the activity itself resulting from contamination or activation.

(i) High-activity wastes: Generally speaking, since it is not easy to assess the activity of solid waste in curies, it is usually expressed as a radiation level (R/h). The limit above which a waste is considered highly active varies from site to site and may be determined by the presence of alpha activity. An important part of highly active solid wastes consists of dismantled equipment and reactor parts.

(a) Collection: The wastes are collected in appropriate containers offering the necessary protection against excessive radiation exposure to personnel. These containers also provide protection against the risk of dispersion of the contamination. Where practical, remote or automatic handling of these materials is desirable to minimize personnel exposure. Provision should also be made for temporary storage of wastes that may be produced in larger than normal volume from time to time.

(b) Treatment: In normal operation these wastes are not subjected to any treatment. In the case of contaminated pieces of equipment or reactor parts, it may be desirable to transfer part or all of the activity to a liquid phase by an appropriate decontamination procedure because of the value of the tools or equipment, or to minimize subsequent handling problems.

24


Highly active solid wastes resulting from the treatment of liquid wastes are usually fixed in a final form by incorporation in concrete or other containment.

(c) Storage: Highly active solid wastes are usually stored in pits or trenches under conditions that guarantee confinement of the activity and might allow eventual later recovery.

(ii) Low-activity wastes: These wastes are usually rather voluminous. Normal practice is to reduce their volume to permit storage. Experience has shown that the activity of this type of waste is of the order of a few tenths of a mCi/t.

(a) Collection: Normal commercially available cans or waterproof bags of paper or plastic are used for collecting this type of waste. These wastes do not represent a risk as far as radiation exposure is concerned, and the risk of dispersion of the activity is minimal. It is usually helpful during collection to segregate these wastes according to the treatment they will receive.

(b) Treatment: Treatment of low-activity wastes is not always strictly necessary; when it is done, it is aimed at reducing the volume and confining the activity. The processes applied are a function of the physical nature of the wastes and can be classified as:

combustible,non-combustible but compressible,non-combustible and non-compressible.

Numerous publications give pertinent information on these treatment processes (see in particular Ref. [7]).

(c) Storage or disposal: Low-activity solid wastes can be stored according to different procedures depending on local conditions. In particular, when sufficient space is available and under favourable geological conditions, it is possible to dispose of these wastes by burying them in the ground. This procedure implies a site control practically unlimited in time, and possible recovery in case of an accident. This storage procedure is dealt with in Ref. [8].

25


Some countries adopt a more conservative approach towards disposal of low-activity solid wastes, and use abandoned mines as storage sites, although these are more appropriate for storing high- activity wastes. Reference [8] gives information on this subject. Low-activity wastes may also be disposed of into the sea when provided with containment appropriate to the radioactivity level. This method is treated in Ref. [9],

The last two procedures (underground storage and disposal into the sea) offer the advantage of freeing the reactor owner from any further surveillance.

II.4.1.2. Liquid wastes. Liquid wastes are much more uniform in nature than solid wastes, and in most cases they are aqueous. Lubricating oil and contaminated solvent present a particular case (see Section (d) below). Generally speaking, the specific activity is of the order of 10"6 to 10-2 AiCi/ml but may be as high as a few (1-5) /nCi/ml in coolant letdown or leakage from light-water re actors or from tritium in the moisture condensed from the gas circuit in UK reactors.

(a) Collection: Collection is normally performed by means of permanent piping systems. In certain cases it may be necessary to make use of portable tanks or casks when small volumes are involved. Care should be taken to ensure that these containers are protected from breakage by impact, freezing, etc.

It is to be noted that a certain hold-up capacity may be required for process operation as well as to provide time for radioactive decay before release. Additional hold-up capacity should also be considered where appropriate for infrequent abnormally large volumes of liquid waste, as from major repairs or decontamination.

(b) Treatment: The choice between different processes is dictated by the chemical composition and the specific activity of the wastes. The processes most often used are :

decantation and filtration,evaporation,ion exchange.

These techniques are analysed in Ref. [7] .

26


In normal practice, decantation and filtration treatment are in many cases sufficient before dilution and release of the waste.

Laundry wastes constitute a special category because of the problem of bio-degradation of the detergents used. Usually they are not treated because the activity level is quite low, but provision for adding equipment to allow treatment might be advisable. Difficulty is often experienced in treating wastes containing detergents; however, experience has shown that such wastes do not normally require treatment before discharge.

The residues from liquid-waste treatment (filter beds, evaporator concentrates, ion exchange resins) are considered as solid wastes.

(c) Release: The treated liquid wastes are released to the environment (river, lake or sea) under conditions of appropriate dilution, which means that the norms recommended by the ICRP for the exposure of the population are observed. It is useful to have a tank system for collection of wastes followed by a system for dispersion. Such an arrangement provides an opportunity for decay and analysis before release.

(d) Lubricating oil and solvents: Lubricating oil and solvents constitute a separate category of waste. Small quantities of these wastes canbe absorbed on a solid material such as vermiculite and further treated as solid waste. For large quantities, incineration and distillation appear to be suitable techniques.

II.4.1.3. Gaseous wastes. The quantities of gaseous wastes produced in a power reactor vary largely with the type of reactor. Nevertheless, in all cases the ventilation air of potentially contaminated areas of the installation is released to the environment after filtration. Used filters constitute solid wastes of low activity and are treated as such.

The activity of the gaseous effluents is very variable. For re actors in France, Canada and the United Kingdom the radionuclides most frequently involved are 41A and tritium. For light-water r e actors the fission gases xenon and krypton will usually need controlling. Tritium maybe produced by fission, activation of deuterium or neutron reaction with lithium present as an impurity in graphite or pH control. Fast neutron reaction with boron in light-water re actors also produces tritium.

27


(a) Collection: The collection of the gaseous effluents is usually performed by permanent pipes or ducts. However, local sources of airborne activity may be conveniently handled by temporary local ventilation facilities.

(b) Treatment: Except for ordinary gas filtration there is normally no other treatment. However, it can under certain circum stances be of importance to apply decay storage before release to the atmosphere. In other circumstances specific filtration can be recommended or even sometimes needed. Specific filtration such as by activated charcoal is frequently used for the retention of 131I.

(c) Release: After decay and/or filtration, the gases are r e leased to the atmosphere through a stack. The height of the stack is calculated to dilute the released activity to levels that conform to ICRP recommended norms. Reference [10] contains practical information on this subject.

II.4.2. Relationship to process costs

Determination of costs is a rather complex question which has been studied extensively and described in Ref. [12].

During the planning stage it is essential to develop a detailed study of the waste problem to enable adequate sizing of the installations. For example, it may be necessary to decide whether to operate a simple installation on a continuous basis rather than a more complete facility on a one-shift basis. In the first case, the investment is limited but the operational expenses are high. The second case presents the inverse situation. Generally the investment costs are kept separate from the operational expenses. Operational expenses include essentially the costs of labour and supervision; materials, utilities and services; maintenance; depreciation and general overheads of the station. Other costs which many sites also include are those for applied research and interest on the capital investment.

The experience accumulated up to now shows that investments in waste-management installations are about 1% of the total investment for the power station. The operational expenses are of the same order but are somewhat higher compared with the total operational expenses for the power plant.

28


n.4.2.1. Solid wastes. The cost of managing solid wastes is principally affected by the treatment process and to a much less extent by the storage or disposal operation.

(a) Treatment: Volume reduction by compaction appears to be the least expensive method of treatment. Incineration is much more expensive, because of the investment required.

If after volume reduction the wastes are conditioned by cdn - creting or any other method, costs are consequently increased.

(b) Storage or disposal: Direct storage or disposal into the ground on the site of the station is the cheapest method. The expenses increase considerably if it is necessary to place the wastes in (concrete) trenches or pits.

If the wastes are removed from the site of the power station to be placed in underground caverns or disposed of into the sea, significant operational and transport costs cam be expected.

11.4.2.2. Liquid wastes. The costs of managing liquid wastes are essentially dependent on the treatment process. The maintenance costs for the collection and release systems constitute a relatively small fraction of the total costs.

Treatment by decantation and filtration is much less expensive than evaporation. The treatment by ion-exchange resins requires a relatively complex installation and rather high maintenance costs.

11.4.2.3. Gaseous wastes. Gaseous wastes are not usually treated in the same sense of the term as applied to liquid wastes. Except in PWRs, gaseous wastes are not "collected" but rather released continuously as they are generated, with filtration (and perhaps a short delay for decay as in BWRs). Thus, the costs associated with them (except for the capital costs of tanks, blowers, ducting, stacks, etc.) are largely operating costs associated with the replacement of filters and other maintenance.

II.5. Performance verification and monitoring

Verification of performance and monitoring are essential. Both operations serve to verify that the workers and the population are not exposed in excess of the norms recommended by the ICRP. This brings up the problem of responsibility, especially concerning

29


the public. All countries have competent authorities responsible for the health and safety of the public as affected by nuclear power stations. In any event, the power station operators will have the greatest interest in recording as precisely as possible all operations performed on the radioactive wastes. In particular, all quantities released to the environment or stored should be recorded. This accounting should be part of the station records, and these records should be held accessible for a period of time in accordance with the legislation of the country. These records should be sufficiently precise and also sufficiently concise, so that their consultation after several years is still possible without much difficulty.

11.5.1. Role of in-plant controls

The role of in-plant controls is primarily to protect the workers. The risks to which the workers are exposed are radiation exposure and contamination (internal and external). However, it is also within the plant that control of off-site radiation exposures must be maintained. That is, proper waste management cannot be accomplished by controls or measurements made in the environment; wastes must be measured and controlled within the plant before release.

11.5.1.1. Radiation exposure risks. These risks are associated with collection, treatment and storage (release) of solid, liquid and gaseous wastes.

(aj Solid wastes; Radiation exposure can occur during collection when highly active wastes are involved or during treatment and storage of low-activity wastes. In the first case, the risk can be coped with by using the appropriate protective equipment and by operation at a distance.

In the second case, adequate operational discipline must avoid uncontrolled accumulation of radioactive material that could produce excessive radiation exposure (e.g. incinerator ashes or evaporator concentrates).

It seems highly unlikely that risks of exposure will appear during storage if they have not been identified during collection and treatment. The best way to avoid these risks is probably by limiting the total activity contained in any storage area and by measuring the radiation throughout the processing.

As far as storage in the ground is concerned, a thorough and permanent control must be performed.

30


(b) Liquid wastes: The only radiation exposure risk during collection of liquid wastes would be due to an unforeseen increase in specific activity. Except in case of incident or accident, this risk is of little importance. Nevertheless, particular attention must be given to specific points in the circuit (bends, valves, decantation pots) where activity may be expected to accumulate. As for the treatment, the exposure risk is the same. It is important to follow attentively the activity of the filters, evaporator concentrates, ion- exchange resins, etc. Besides the immediate danger, a too large accumulated activity could constitute a problem during the treatment of these residues. Given the low specific activity of the liquid wastes after treatment, there will be no exposure risks asspciated with the release operation.

(c) Gaseous wastes: The case of the gaseous wastes is almost the same as that of the liquid wastes, except for abnormal increase of the specific activity. Exposure risk can only exist in case of significant and uncontrolled accumulation of activity on the filters and during successive decay storage. In principle, radiation exposure risks are very low.

II.5.1.2-. Contamination risks. Contamination risks are due to dispersion of the activity in the atmosphere.

(a) Solid wastes: Dispersion of activity can take place during collection if the containment has not been executed carefully. During treatment it can also occur in case of equipment failure or malopera- tion (during removal of incinerator ashes).

In any case, the risk will be proportional to the activity involved. It will also depend greatly on the physical and chemical nature of the waste, products in powder form being the most dangerous.

(b) Liquid and gaseous wastes: Dispersion of activity can only take place in the event of leakage from the piping and storage tanks or during treatment. Preventive maintenance will be necessary to reduce this risk to a minimum. Precautionary measures can also help to reduce such risks by providing monitored sumps and surface protection, i.e. plastic film or absorbent paper.

31


11.5.2. Role of environmental monitoring

The role of environmental monitoring is to ensure that the public is not exposed to radiation doses above the recommended ICRP norms. Even if the power plant owner or operator does not assume the legal responsibility of the surveillance of the public, he should perform environmental monitoring and record the results of the measurements to be able to compare them with those of the responsible authority.

It should be pointed out that if in-plant control is performed adequately, environmental monitoring could be reduced correspondingly to a minimum. In particular, if no incidents occur in the power station it is highly improbable that an abnormal situation will declare itself in the environment.

In general, environmental surveillance will have to be adapted to local conditions. For example, radioiodine is among the most radiotoxic of the readily volatilized fission products, yet it is not so essential to impose a special effort on the search and determination of 131I in a region without grazing cattle or when the release of radioiodine to the environment is known to be inconsequential. Similarly, the existence of a significant shellfish bed requires that shellfish be included in the surveillance because of their significant reconcentration of certain radionuclides. Finally, it should be noted that sampling methods and analytical techniques and instruments must be appropriate for the radionuclides present. Tritium, for example, requires somewhat special consideration.

There is no problem of environmental monitoring for solid wastes since they are stored under appropriate conditions on the site itself or at another specially selected place.

11.5.2.2. Liquid wastes. It may be important to monitor the body of water in which the liquid wastes are released and diluted. It can be of particular interest to study concentration phenomena and it is desirable to perform systematic analyses of the water down-stream from the release point. It can be equally profitable to monitor the activity fixed by the flora and fauna of the river, the lake or the sea (in particular, fish) or crops irrigated by water into which liquid wastes are released. In many cases it will also be useful to control the ground water by sampling the water in pits or bydrilling.

32


II.5.2.3. Gaseous wastes. Deposition of fine particulates can occur at rather distant points from the release point and thus affect im portant areas. Environmental monitoring is all the more sensitive because of this. The low activity levels involved make it necessary to perform a sufficient number of measurements to allow a statistical precision appropriate to the circumstances. The measurement points will have to be carefully distributed and, in particular, the monitoring will have to be performed with special care in the dominant wind direction. Verification should bear upon vegetation and crops, and in some special cases detailed analyses at the time of harvesting of crops.

33


FUTURE TRENDS AND PERSPECTIVES (Section III)


III. FUTURE TRENDS AND PERSPECTIVES

The persistent nature of many of the radionuclides likely to appear in wastes from nuclear power stations suggests that some advance consideration of the future aspects of such facilities is desirable. It is, of course, both difficult and dangerous to attempt to predict the future. Nevertheless, past experience suggests that the exercise may be of some use to those groups entering on nuclear- generating programs. Some of the considerations are somewhat repetitive of those expressed in the previous sections, but they are worth repeating for emphasis.

It is important to recognize again the excellent performance of waste-management facilities at existing nuclear power stations. In no case has there occurred any exposure in the plant environs which has even approached, let alone exceeded, the recommendations of the ICRP or national regulations. This has been accomplished at a very small fractional cost in capital investment and in operating charges irrespective of reactor type. Further, there is no present reason to expect any deterioration in this performance in the future. However, some considerations that may affect the performance requirements, and hence costs, are presented below.

At the outset of a nuclear power program, the plant operator and the regulatory authorities need to consider only single stations, probably located at a considerable distance from large population groups. As the decisions are made to install additional nuclear generating capacity, these additional plants may be sited sufficiently close to the existing stations to produce interaction on some common aspects of their environment.

Alternatively, the original station may be supplemented by additional units at the same site. Both these situations have already occurred in the United States of America, the United Kingdom and France, and are likely to occur elsewhere.

Some consideration in the design of waste-management facilities is thus required to ensure that the original station has not pre-empted an undue portion of the environmental capacity, or that the newer stations are not restricted to costly waste-management procedure.

Further development of the nuclear-generating program will undoubtedly create pressures by the operating organizations to site their stations as close to centres of major power demand as feasible in order to minimize the costs associated with transmission of power. In some cases, this practice may be mandatory if the station is used

37


to generate steam for heating or process industry. Dual-purpose plants of the latter type are currently under consideration in the United States of America for electricity generation and desalination of sea water in the Los Angeles area, and for the production of both process steam for a chemical plant and electricity in Michigan.

In such cases, where it is likely that the station will be in proximity to larger population groups, special consideration Will be required in the design and operation of waste-management facilitie's. These considerations may include more effective and elaborate waste- control processing and monitoring schemes, changes in philosophy in regard to on-site waste accumulation, elaborate transport facilities 1 for off-site disposal of waste materials, etc. The safety implications and costs of these and.other concomitants of siting power stations in urban areas must also be included in considering overall benefits and risks of such stations.

There are other areas to which consideration must be given as a nuclear power program begins to develop momentum, only some of which can be seen at present. One of these is concerned with the accumulation of such longer-lived radioactive materials in the environment as tritium and 85Kr. Although the magnitude of release of these materials from power reactors is strongly dependent upon the particular reactor design, they can ultimately be released in any reactor cycle that includes reprocessing of spent fuel. According to the data reported in the Appendixes, releases of tritium from reactors in Canada, the United Kingdom and the United States of America have been much larger in amount than those of other nuclides,' although it should be stressed that the relative toxicity of tritium is quite low in comparison with that of other radionuclides. It appears, however, that in view of the present lack of feasible methods for extraction of tritium from nuclear-power-station effluents, the accumulation of this nuclide in the environment may be worthy of special consideration in some locations when large numbers of nuclear stations are operating some decades hence.

85Kr does not appear to be of particular significance in power- station waste management, as indicated in the Appendixes. Although this nuclide is released in spent fuel reprocessing, a promising low- temperature adsorption process, which is moderately costly, but effective, may permit future recovery if'necessary.

Another waste source, which is not normally considered but which will ultimately be produced at any power station, is that resulting from the decommissioning of the station at the end of its useful

38


life. The normal economic lifetime of a power station is in the range of 30-50 years, although developmental or prototype power stations may have shorter lifetimes. Since the shutting down and disassembly of a large nuclear station may entail considerations of waste management not normally provided for during the operating lifetime, advance planning for this ultimate requirement would be most useful. Some benefit from this type of experience can be obtained from the shutdown and disassembly in the United States of America of the Hallam Sodium-Graphite and the Piqua Organic Moderated Reactors; and (perhaps less appropriately) from the decommissioning of the Wind- scale reactor in the United Kingdom, the NRX reactor in Canada and the SL-1 reactor in the United States of America following their respective accidents.

In line with the foregoing it would be appropriate to note at this point some comments on the question of ultimate disposition of wastes from nuclear stations. As indicated in the Appendixes and in other publications, national practices vary in some respects in regard to this question. At all sites some controlled disposals of liquids and gases are made to the environment. Methods for determining limits for discharge of fluid wastes vary somewhat between nations, although all are based on the recommendations of the ICRP in regard to radiation dose limits. In France and the United States of America, control is usually established on the basis of concentration limits derived from the ICRP; an exception may be noted for gaseous releases from boiling-water reactors where an annual average release rate is specified because of the relationship between quantity (not concentration) released, and the ambient concentration or exposure at a particular location. In Canada and the United Kingdom, controls imposed on effluent disposals generally specify radioactivity release rates, rather than fluid concentrations. The net effect of either practice is identical, however, in assuring compliance with appropriate dose limitations.

Waste-management practices for solids are somewhat less similar. In Canada, disposal by burial at the plant site has been practised; in the United States of America this is prohibited, and solid wastes are temporarily stored for shipment to off-site burial grounds on land owned by a State or the Federal Government; current French practice is to store solids on-site, although this is not the ultimate disposal method to be followed; the United Kingdom follows a somewhat similar method, although the engineering approach differs by incorporating in the station facilities (e .g . vaults) designed to

39


accumulate and retain the waste solids in recoverable form for the lifetime of the station, or longer if necessary.

Each of these practices is considered acceptable to maintain exposures to the public or plant employees at a minimum. However, each scheme considers that the question of ultimate disposal of these wastes is worthy of additional study to minimize the impact of this waste-management process on the prime purpose of nuclear-station operation, i. e. the generation of electricity.

In looking at future prospects for waste management, there would appear to be varying degrees of incentive for change. The success to date in the control of wastes would imply little reason for modification except perhaps in the areas of solid-waste disposal discussed above. Such changes as are likely to occur may be expected in improved fluid processing and concentration techniques for solid and liquid wastes. However, these changes are likely to be relatively minor engineering modifications of present systems intended to improve their safety, efficiency or ease of operation, rather than basic changes in unit operations. Examples might include improved provisions for the incineration of low-activity combustible wastes as at UK installations; and the proposed substitution of finely powdered ion-exchange resins for separate precoat filters and deep-bed demineralizers at new US boiling-water reactor stations, etc.

This prediction is not intended to discourage investigations of new techniques for waste-management systems which will improve the effectiveness and lower the costs of these systems still further. It is intended merely to suggest the view that newly developing nuclear-power programs need not expect major changes in concept or costs for these systems from those estimated in the IAEA Technical Reports Series No. 83 [12] and in the Appendixes.

Finally, it is appropriate to consider the future in respect to the types of nuclear power stations currently in the early stages of development, such as the breeder reactors. Little has been said of these because of the present lack of data. However, new and different technical problems can surely be expected to accompany these new station designs. These will probably include the difficulties of dealing with liquid-metal-coolant wastes, as well as others at present unforeseeable. It can also be expected that the first power stations of a given design will include rather more elaborate and conservatively established waste-management systems to ensure that adequate plant margin exists for handling unexpected waste generation. It is believed that, provided comprehensive examina

40


tions and analyses are made of both station systems and characteristics, and of the plant environment, and provided adequate margins of safety are included in design and operation of waste- management systems for these plants, the performance of the new stations can be expected to conform to the excellent example provided by those at present operating.

41


TABLES I-IV (Section IV)


TABLE I. WASTE MANAGEMENT DATA FROM OPERATING CANADIAN POWER REACTORS (MARCH 1968)

NPD Douglas Point

Power (M W (c)) 25 200

Operating period 11 Apr. 1962 - 7 Jan. 1968 -

Gross e le ctr ica l MWh 750 000 250 000

Cladding Z irca lloy 2 Z irca lloy 2

D efective fuel < 0 .0 1% < 0 .1 %

Gaseous waste

Treatment provided Decay storage Decay storage

Treatment used nil nil

Stack exhaust rate (m 3/s ) 11 28

Stack height (m ) 50 46

Permissible average release:

N oble gases (C i/d ) 3 X 10s 4X 10 4

Halogens (C i/d ) 0 .1 C i 0 .2 5 Ci

Tritium (C i/d ) 4X 105 2 X 104

Range o f average release:

A ctivation and noble gases (C i/d ) <10 12a

Vo o f lim it 0 .3 0 .03

Halogens J iC i/d ) < 5 0a

% o f lim it <5 X 10_> 0

Tritium (C i/d ) <300 3a

% o f lim it <8 < 0 .0 2

44


TABLE I (cont.)

NPD Douglas Point

Liquid waste

Treatment provided "1

Treatment used J

Ion exchange on fuel storage bay ^

N eutralization , concreting

None

Volum e co lle cte d (10s 1/yr) 1900 4000

% returned to reactor use nilc nilc

A ctiv ity range before dilution (jiC i/m l) io"5 - i o ' 6 6 X lo ’ 7 -1 X 1 0 "5 a

Condenser water for dilution(1 /s ) 1 0 0 - 300d 15 000

A ctiv ity discharged:

Total (C i/y r) 0 .18 2 .0 *

Concentration range (fiC i/m l) 10"*-4 X 10‘ 7 4X 10‘ 9 -3 X 10'8a

Tritium (C i/y r) 3700 0a

°lo o f overall lim it 6 0

Solid waste

Annual average amount 25 000 kg 1000 kg 8 m3a

Annual average activity < 200 mR/h > 200 mR/h 20 C i a

Resins

Annual average amount (kg) 10 104a

Annual average activity > lR /h 3 .5 C ia

E stim ated from 3 months operating experience, b Applies to both reactors.c Part o f the fuel storage bay water is continuously recirculated through ion exchangers, and

part is bled to waste to keep tritium concentration at an acceptable leve l. dThis is "Process Water" — variable amounts o f condenser water are added.

45


< Ji T A B L E II. RESUM E O F D A T A A N D O P E R A T IO N A L E X P E R IE N C E IN FR A N C E

G .2 G . 3 E d F l EdF 2 EdF3

Gross power 250 MW 250 MW 300 MW 850 MW 1 50 0 MW

Power M W (e) 4 0 MW 40 MW 70 MW 180 MW 475 MW

O perating period April 1959 - D ecem ber 1967

April 1960 - D ecem ber 1967

June 1963 - D ecem ber 1967

February 1965 - D ecem ber 1967

O ctober 1967 - D ecem b er 1967

Gross e le c t r ic a l output MWh (m inus au x iliaries) 1 708 528 1 5 7 3 778 8 17 000 1 9 0 6 000 115 000

C ladding M agnesium + 0 .7 °}o z irconiu m

M agnesium +0 .7 % zirconiu m

M agnesium allo y M agnesium allo y M agnesium alloy

M axim um number o f fu el assem blies with d e fe c tiv e cladding

Not ap p licab le s ruptured fu e l slugs are in any ca se im m ed ia te ly discharged. T h e proportion is very low - in th e 1 0 " 5 range.

G aseous w aste T h e only gaseous wastes are from th e co o lan t ( C 0 2) , w hich, when th e system must b e em p tied , is re leased into th e atm osphere a fter filtra tio n .

T rea tm en t provided R ecy clin g , in c e r ta in cases d ischarge

R ecy clin g , in certa in cases discharge

S y stem atic discharge S y stem atic d ischarge Sy stem a tic d ischarge

T re a tm e n t used ditto ditto ditto ditto ditto

S ta c k exhaust r a te (m 3/h )

5000 5000

S ta c k h eig h t (m ) ab ov e ground

45 45 40 30 -

P erm issib le average Not ap p licab le ; th ere is no re lease T h ere are no d ifficu ltie s with the rare re leases; th e only rad io activ e gas

ann ual re lease

Range o f av erage annual re le a se

• is 41A , from ac tiv a ted argon present in th e C 0 2 only as an im purity


T A B L E II (con t.)

Liquid w aste

T rea tm en t provided

T rea tm en t used

V olum e c o lle c te d (10s m s/y r)

P ercen tag e returned to reacto r use

A ctiv ity range before dilu tion

A ctiv ity discharged, to ta l per year in cu ries

C on cen tration range

T ritiu m (estim ated )

Production o f o il

So lid w aste

Annual a v erag e, v o lu m e and activ ity

Resins

Others

T rea tm en t provided

T rea tm en t used

F iltra tio n

F iltra tio n

Filtra tio n

F iltra tio n

M axim um 10 for th e two reactors

Not ap p licab le | Not ap p licab le

About IQ-5 C i/m 3

IConsiderably lower than 1 C i

About 10"® C i/m 3

n il | nil

From 1 to 2 m3/y r (on average)

n i l nil

140 m 3 for the two reactors

Com pression and cem en tin g ; in cin eration ; chopping o f m e ta llic w astes

ditto

Evaporation

n il

Not ap p licab le

nil

Evaporation

n il

Evaporation

nil

About 100 000 m 3 for th e th ree reactors

| Not ap p licab le j Not ap p licab le

From 1 0 '5 to 1 0 '1 C i/m 3

I I1 . 4 C i for th e th ree reactors

< 10 ‘ 5 C i/m 3

| n il

5 m3/y r for th e th ree reactors

n il

200 ms for th e th ree reactors (m ost o f it in 5 0 0 - l i t r e drums; th e a c tiv ity at co n ta c t being low er than 500 m R /h )

Com pression

ditto


T A B L E III. W ASTE M A N AG EM EN T D A T A F R O M O P E R A T IN G POW ER R EACTORS

Station Berkeley Bradwell Hunterston "A "

1 . Power rating MW (th) 1116 1062 1150

M W (e) J 276 300 300

2 . O p eration al period reported (S e e also item 11)

June 1962 - D ecem ber 1967

July 1962 - D ecem ber 1967

February 196 4 - D ecem ber 1967

3 . Gross e le c t r ic i ty gen erated (kW h) x 10s

10 606 12 628 8 500

4 . F u e l claddin g m ateria l Magnox M agnox M agnox

5 . M axim um number o f fu e l e lem ents "b u rst"/y r W

This inform ation

G aseous wastes 6 . T rea tm en t provided Filtration F iltra tio n F iltra tio n

7 . T rea tm en t used F iltra tio n F iltra tio n F iltra tio n

8 . T o ta l d ischarge ra te W ( f t3/m in ) (per reactor)

1 0 0 8 0 0 2 1 6 0 0 0 208 000

9 . No. o f stacks in use (per reactor) 2 2 5

S ta c k height (f t) above ground le v e l 175 137 200

1 0 . P erm issib le annual ac tiv ity re lease G aseous ") P articu la te J A ll authorizations to d ischarge rad io activ e gases, m ists and dusts

1 1 . Range o f annual av erage activ ity r e le a se ^

G aseous (m C i/s ) 5 .5 7 .8 Not av a ilab le

P articu la te (p C i/d )( a ) + (b) 27 - 39

(a)1 1 -1 3

(b)0 . 0 7 - 0 . 4

Not av a ilab le

O p eration al period 196 5 - 1967 1966 1967

NOTES1 . A ll reactors h av e Burst C artridge D etectio n (BCD) system s th at o p erate con tin u ally ; when a "burst" is d etected the

e lem en t is rem oved from th e reacto r and is p laced in a closed m e ta l container or "b o ttle " for storage in th e fu el e lem en t coo lin g pond. In th is way the re lease o f activ ity from fu el e lem en t can bursts is m in im ized .

2 . Item 8 refers to sh ie ld -co o lin g a ir discharges o n ly . D ischarges and losses o f CO j from the reactor and an cilla ry pressure c ircu its a re sm all by com parison, am ounting in norm al operation to one or two tons o f C 0 2 per reactor per day. D ischarge or "blow dow n" o f th e m ain pressure c ircu it is m ade only o n ce or tw ice per y ear , taking about 4 h to co m p le te .

3 . No com prehensive and re lia b le data on m easured gaseous activ ity discharges are a v a ila b le . T h e m ajor com ponent o f gaseous activ ity discharged from s te e l pressure vessel reactors, how ever, is 41A in sh ie ld -co o lin g a ir , design estim ates for w hich are quoted in item 11 .

48


IN THE U N ITED KINGDOM

Hinkley Point Trawsfynydd Dungeness "A " S iz e w ell Oldbury

1942 1720 1680 1896 1784

500 500 550 580 600

M arch 1965 - D ecem ber 1967

• M arch 1965 - D ecem b er 1967

O ctober 1965 - D ecem ber 1967

M arch 1966 - D ecem ber 1967

N ovem ber 1967 D ecem ber 1967

1 1 0 0 9 6 3 6 1 7 654 4 4 6 8 95

Magnox M agnox Magnox Magnox Magnox

on fu e l e lem en ts is not ex tra cted from th e op eratio n al records

F iltra tio n Filtra tio n Filtration Filtration Filtration

Filtra tio n Filtration F iltration Filtration Filtration

2 0 4 000 350 4 00 2 8 0 0 0 0 218 000 Nil

1 2 2 1 -

180 206 169 192 -

require th at th e best p ra c tica b le m eans sh all be used to m in im ize the am ount o f rad io activ ity discharged

1 2 .2 7 .5 8 . 4 7 .2 N il

(a) (b) (a ) + (b) (a) (b) (a) (b) (b)6 4 -1 0 4 0 . 3 - 3 . 1 3 1 - 70 1 0 2 - 112 4 0 -4 0 0 4 -3 3 6 *9 0 . 4

1966 ■ 1967 1966 - 1967 1966 1967 1966 - 1967 1967

In item 11 , (a ) = p a rticu la te in sh ie ld -co o lin g a ir , and (b ) = p articu la te in C 0 2 . T h ese discharges are shown com bined or separately depending on th e m onitoring arrangem ents at th e particu lar station . During 1966 CEGB stations adopted a revised m ethod o f assessing and reporting the p articu la te rad io activ ity discharged in gaseous wastes (th e revised m ethod is based on sam ple counting after natural rad ioactiv ity has d ecayed aw ay). For this reason th e operational period reported for p a n ic u la te activ ity discharged differs from th at quoted in item 2.

4 . Ion ex ch an ge em ployed sin ce 196 7 .5 . Excluding tritiu m .6 . D ischarges are m ade to a r tif ic ia l fresh water la k e .7 . No com prehensive and re lia b le estim ates o f solid waste rad ioactiv ity co n ten t are a v a ila b le .8 . G iap h ite s leeves, support m em bers and zircon iu m bars in cluded .

49


T A B L E III (con t.)

Station Berkeley Bradwell Hunterston "A "

Liquid wastes12. T rea tm en t a v a ila b le

13. T rea tm en t used

F iltra tio n and ion

(4)ex ch an g e ' '

F iltra tio n and ion exchange

Precip itation andsand

filtra tio n

14. A verage vo lu m e c o lle c te d annually (g a l/y r )

2 .4 X 10® 1.2 x 10® 1 .6 x l O 7

15. P ercen tag e re-used N il Nil N il

16. A ctiv ity r a n g e ^ before dilu tion in C . W. system ( jiC i/m l)

1 0 " 7- 2 .1 0 " * 1 0 '1 - 2 . 1 0 '15 x 10"s 7 X 10‘ 4

17. Estim ated activ ity r a n g e d after d ilu tion in C . W. system ( jiC i/m l)

4 . 1 0 ' " - 9 . 1 0 " 6 3 . io ‘ “ - e . i o -55 x 1 0 '9 7 x 10*®

18. F lo w -ra te o f coo lin g water a v a ila b le under norm al operating conditions for im m ed ia te dilu tion o f wastes (g a l/m in )

3 .5 X 10s 1 .5 X 10s 2 .6 4 X 10s

19. M axim um p erm issib le annual activ ity d ischarge (C i/y r)(a) A ctiv ity other than tritiu m 200 200 (including

* 5 C i o f 65Zn

200

(b ) T ritiu m 1500 1500 1200

20. A verage annual ac tiv ity discharge (C i/y r)(a ) A ctiv ity other than tritiu m 1 8 .0 2 7 .2 10

(b ) T ritiu m 358 3 31 275

S o lid w a s t e s ^2 1 . A verage annual volum e ( f t 3)

F u e l e lem en t debris W et m ateria ls ( f ilte r sludges, sands

and resins)So lid com bu stib le wastes (in cin erab le ) Other wastes

2 700

64 10 2 00

2800

440

16 12 0 00

9 000^ )

N il35002000

NOTES

1 . A ll reactors have Burst C artridge D etectio n (BCD) system s th at o p erate co n tin u a lly ; when a "burst" is d etected the e lem en t is rem oved from th e reactor and is p laced in a closed m e ta l co n tain er or "b o ttle " for storage in th e fu el e lem en t coo lin g pond. In th is way th e re lea se o f a c tiv ity from fu e l e le m e n t ca n bursts is m in im ized .

2 . Item 8 refers to sh ie ld -co o lin g air discharges o n ly . D ischarges and losses o f COt from th e reactor and an cilla ry pressure c ircu its are sm all by com parison, am ounting in norm al operatio n to on e or two tons o f CO* per reacto r per day. D ischarge or "blow dow n" o f the m ain pressure c ircu it is m ade only o n ce or tw ice per y ea r , taking about 4 h to co m p le te .

3 . No com preh ensive and re lia b le data on m easured gaseous activ ity discharges are a v a ila b le . T h e m ajor com ponent o f gaseous activ ity discharged from stee l pressure vessel reactors, how ever, is 41A in sh ie ld -co o lin g a ir , design estim ates for which are quoted in ite m 11 .

50


H inkley Point Trawsfynydd Dungeaess "A ” SLzew ell Oldbury

Filtration F iltra tio n and ion ex ch an ge

Filtra tio n F iltra tio n F iltra tio n '

3 .6 x 1 0 s 3 . 0X 10 ‘ 1 .9 x 106 2 .4 X 107 2 .2 X 1 0 6

Nil Nil Nil Nil N il

1 0 '8 - 2 . 10"3 10 -5 - 4 . 10-2 8 . 10~6- 6 . 10~Z 1 . 1 0 'S- 4 .1 0 " 8 1 . 10*4 -1 .1 0 "S

4 . 10'12 - 7 . 10*7 10 •’ - 4 . 10"‘ 4 . 10’ 9 - 3 . 10” 5 8 . 10"9- 3 . l ( f 6 3 . 10’ 8-3 . lO-7

5 .8 x 105 2 .9 X 10s 3 .5 X 10s 2 .3 X 105 3 . 5X 105

200 40<6> 200 200 100

2000 2000 2000 3000 2000

9 .8 3 .3 1 5 .4 1 .9 0 .9

72 .7 118 401 79 .1 2 2 .2

214 54 . . .

N il 250 _ N il .20 000 33 000 3800 13 000 9400

160 77 840 270 -

In ite m 1 1 , (a ) = p a rticu la te in sh ie ld -co o lin g a ii, and (b ) = p articu la te in C 0 2 . T h ese discharges are shown com bined or sep arately depending on th e m onitoring arrangem ents at th e p articu lar station . During 1966 CEGB stations adopted a revised m ethod o f assessing and reporting th e p a rticu la te rad ioactiv ity discharged in gaseous wastes (th e revised m ethod is based on sam ple counting after natural rad io activ ity has decayed aw ay). For this reason th e operatio nal period reported for p ar ticu la te activ ity discharged differs from th at quoted in ite m 2 .

4 . Ion ex ch an g e em ployed sin ce 1 9 6 7 .5 . Excluding tritiu m .

6 . D ischarges are m ade to a r tif ic ia l fresh water la k e .

7 . No com prehensive and r e lia b le estim ates o f solid w aste rad ioactiv ity content a re a v a ila b le .8 . G raph ite sleeves, support m em bers and zirconiu m bars in cluded .

51


T A B L E IV . SU M M A RY O F U N ITED ST A T E S DESIGNS A N D O PE R A TIN G E X P E R IE N C E

Boiling water reactors

D resden-l Big Rock Point Hum boldt Bay

Power rating 700 MYV(th), 2 0 0 MW(e) 157 M W (th), 50 MW(e) 165 M W (th), 52 MW(e)

O peration al period reported •Oct. 1959 - D e c . 1966 Sep . 1 962 - A pril 1967 F e b . 1963 -F e b . 1967

Gross e le c t r ic i ty gen erated (MWh) 6 6 00 0 00 1 0 5 3 0 00 1 0 5 5 0 00

Fu el cladding m ateria l Sta in less s te e l, Z ircaloy Stain less stee l S tain less s te e l, Z ircaloy

M ax . assem blies w ith d efectiv e cladding {%

5 15 25

G aseous wastes:

T rea tm en t a v a ila b le 2 0 -m in d elay , filtra tio n 3 0 -m in d elay , filtra tio n 1 8 -m in d elay , filtra tio n

Trea tm en t used Sam e Sam e Sam e

S ta c k exhaust ra te (c fm ) 4 4 0 00 30 000 12 000

Height o f s ta ck (ft) 300 240 250

Perm issib le annual average

R elease3

A ctiv ation and noble gases 700 0 00 fiC i/s 0 .1 0 6 p C i/s 50 0 00 fiC i/s

Halogens and particu lates 3 .6 MCi/sd 0 .1 8 MCi/sd

Range o f annual average re lease

A ctivation and no ble gases < 100 to 25 000 yCi/s < 2 0 to 35 000 /iC i/s 40 to 1 4 1 0 0 vC i/ s

P ercent o f lim it < 0 . 0 2 - 3 . 6 < 0 . 0 0 2 - 3 . 5 0 .0 8 -2 8

Halogens and particu lates 0 .0 0 2 - 0 .0 0 3 pC i/s < 1 . 2 f i C i / s 1 0 * 5- 0 . 07 jiC i/s

P ercent o f lim it < 3 0 1 -3 8

Liquid wastes:

T rea tm en t a v a ila b le D elay , filtra tio n , evaporation , dem ineralization

D elay , filtra tio n , evaporation , dem in eralization

D elay , f iltra tio n , evapo ration , dem in eralization

T rea tm en t used D elay , filtra tio n , dem in eralization

D elay , filtra tio n , dem ineralization

D elay , filtra tio n dem in eralization

^D erived from lim its stipulated in AEC lic e n c e s .Based on an MPC for noble gas m ixtures o f 3 x 10*8 fiC i/c m 5 .

c Based on an MPC for noble gas m ixtures o f 3 * 10"T /iC i/c m 3 , a con cen tration lim it at poin t o f d ischarge o f 1000 x MPC,and a stack exhaust rate of 15 0 00 ft* /m in (ST P )

^Based on an MPC for halogens and p articu la tes o f 3 x 1CT10 f iC i/cm 3 .e Based on an MPC for halogens and particu lates o f 1 x io * 10 jiC i/c m 8 .

52


Pressurized water reactors

Indian Point-1 Y ankee San Onofre C on n ecticu t Yankee

585 M W (th), 163 M W (e)

Aug. 1 9 6 2 -S e p .1966

3 4 8 9 000

Sta in less s te e l

~ 0

600 M W (th), 163 MW(e)

Ja n . 1 9 6 1 -D e c . 1966

6 3 6 2 0 0 0

Sta in less stee l

~ o

1347 M W (th), 450 MW(e)

Jun. 1 9 6 7 -D e c . 1967

386 500

Sta in less stee l .

~ 0

1473 M W (th), 490 MW(e)

Ju l. 1 9 6 7 -M a r . 1968

1 1 5 7 3 00

Sta in less stee l

~ 0

120-d ay s d elay , filtra tio n

F iltra tio n

2 8 0 0 0 0

400

6 0-d ay s d elay , filtra tio n

Sam e

15 000

150

D elay , filtra tio n

F iltra tio n

4 0 0 00

140

D elay , filtra tio n

Sam e

70 000

175

50 0 00 j jC i/s b

0 . 2 4 p C i/se

2000 fiC i/sc 54 0 00 pC i/s

1 8 .3 p C i/s

10 0 00 133X e eq u iv .) p C i/s

0 .0 7 to 1 .6 /iC i/s

0 .0 0 0 1 3 -0 .0 0 2 6

~ 2 x 10’ 8 MCi/s

<1 0 " 5

0 .0 0 2 -2 2 C i/y r

0 . 0 0 1 - 0 .0 3

1 .2 x 10’ 8 M Ci/cm S

0 .0 0 4

0 - 2 . 3X 1 0 ‘ T pC i/s

2 x lO ’ 9

D elay , f iltra tio n , evapo

ration , d em in era liza tio n , gas stripping

D elay , filt ra t io n , d em ine ra liz a tio n , gas stripping

D elay , f iltra tio n , evapora tio n , d em in eralization

Sam e

D elay , d em in eralization , gas stripping

Sam e

D elay , filtra tio n , evaporation , dem in eralization

Sam e

^Not including tr itiu m .l L im it is based on a continuous d ischarge o f tritiu m (averag ed over 1 2 co n secu tiv e months) o f 3 x lC f5 /iC i/c m 8 .JSo lid w aste a c tiv itie s are "b est estim a tes" a t the tim e o f sh ipm ent. They vary on a "u n it v o lu m e" basis over m any orders of m agnitude.

^ T h is w aste is m ainly evaporator bottom s and in cinerator ashes m ixed with c e m e n t, but also includes resins.* P artia l year.

53


T A B L E IV (c o n t .)

Boiling w ater reactors

D resden-I Big Rock Point Hum boldt Bay

V olu m e c o lle c te d , 103 g a l/y r 1 1 0 0 0 ' ~ 1 0 0 0 ~ 7 0 0

P ercent returned to reactor use 84 - 8 0 0

A ctiv ity range before dilution*( t iC i /c m 3) 10-S - 1 0 '2 1 0 "4 -1 0 " ! 1 0 ‘ 4 -1 0 * 2

Condenser water for dilu tion (gpm ) 167 000 50 0 00 10 000

A ctiv ity discharged

T o ta l C i/y r* 4 . 1 5 .8 1 .3

C oncen tration range* ( C i/c m 3) 1 0 ' * - 5 x 1 0"8 2 x 1 0 ’ 8 -1 0 " 7 2 x 1 0 *9- 1 x 10‘ *

T ritiu m (C i/y r) (estim ated ) 5 -1 0 20 20

Percent o f lim it (estim ated )1

So lid wastes:

< 1 0 "3 0 .0 0 7 0 .0 0 3 5

A verage annual volum e and a c tiv ity 1

Resins 7 70 f t8, i C i 2 5 0 f t3 , 120 C i 50 f t3 , ? C i

Other 3 20 0 f t8 , 100 C i 250 f t s , 100 C i 1000 ft3 , 250 C i

^D erived from lim its stipulated in AEC lic e n c e s .Based on an MPC for noble gas m ixtures o f 3 x 10*® p C i/c m 8 .

c Based on an MPC for noble gas m ixtures o f 3 x 10*1 fiC i/c m s , a co n cen tra tio n lim it a t point o f d isch arg eof 1000 x MPC, and a sta ck exhaust ra te o f 15 0 00 f t s/m in (ST P )

^Based on an MPC for halogens and p articu la tes o f 3 x 1CT10 p C i/cm 3 . e Based on an MPC for halogens and p articu lates o f l x 10*10 fiC i/cm 3 .

54


Pressurized water reactors

Indian P o in t-I Y ankee San Onofre C onn ecticu t Yankee

8 800 850 NA; Volum e released 758 NA: V olum e released 1150

0 0

3 x 10‘ s- 2 x 1 0 "s 10‘ 6 -1 0 " 5 1 .7 X 1 0 ‘ 6 - 4 .6 X 1 ( T 4 5 .5 X 1 0 " 5- 5 .8 x 10 ‘ 4

160 000 140 000 3 5 0 0 0 0 372 000

1 1 .1 0 0 .3 1 7 1 0 . 79*

3 X 10*10-2 X 1 0 " 7 3 X 10*10- 2 X 10 8 x 10‘10 9 x 10‘10 (av . )

500 1300 830

0 .0 3 0 .1 6 10-2

?

900 ft3 , 1 C i

1

360 0 ft3 , 500 C ik

?

?

?

?

*Not including tritiu m .^Lim it is based on a continuous d ischarge o f tritiu m (averag ed over 12 con secu tive months) o f 3 x 10"s /iC i/c m 3 .JSolid w aste a c tiv itie s are "b est estim a tes" at th e tim e o f sh ipm ent. They vary on a "u n it v o lu m e" basis over many orders o f m agnitu de.

^ T h is w aste is m ainly evaporator bottom s and in cin erator ashes m ixed with ce m e n t, but also includes resins.* P artia l year.

55


R E F E R E N C E S

[1 ] INTERNATIONAL COMMISSION ON RADIOLOGICAL PROTECTION, Recommendations o f the International Commission on Radiological Protection, ICRP Publication 9, Pergamon Press, Oxford (1966).

[2 ] INTERNATIONAL COMMISSION ON RADIOLOGICAL PROTECTION, Recommendations o f the International Commission on Radiological Protection, Report o f Committee 2 on Permissible Dose for Internal Radiation, ICRP Publication 2, Pergamon Press, Oxford (1959).

[3 ] INTERNATIONAL COMMISSION ON RADIOLOGICAL PROTECTION, Recommendations o f the International Commission on Radiological Protection, as amended 1959 and revised 1962, ICRP Publication 6, Pergamon Press, Oxford (1964).

[4 ] INTERNATIONAL ATOMIC ENERGY AGENCY, Basic Safety Standards for Radiation Protection, 1967 Edition, Safety Series No. 9, IAEA, Vienna (1967).

[5 ] WORLD HEALTH ORGANIZATION, International Standards for Drinking Water (Report o f an Expert Comm ittee) (1963).

[6 ] INTERNATIONAL COMMISSION ON RADIOLOGICAL PROTECTION, Principles o f Environmental Monitoring related to the Handling o f Radioactive Materials, ICRP Publication 7, Pergamon Press, Oxford (1966).

[7 ] INTERNATIONAL ATOMIC ENERGY AGENCY, Basic Factors for the Treatment and Disposal o f Radioactive Wastes, Safety Series No. 24, IAEA, Vienna (1967).

[8 ] INTERNATIONAL ATOMIC ENERGY AGENCY, Radioactive Waste Disposal into the Ground, Safety Series No. 15, IAEA, Vienna (1965).

[9 ] INTERNATIONAL ATOMIC ENERGY AGENCY, Radioactive Waste Disposal into the Sea, Safety Series No. 5, IAEA, Vienna (1961) 53.

* [10 ] INTERNATIONAL ATOMIC ENERGY AGENCY, Techniques for Controlling Air d ilu tion fromthe Operation o f Nuclear Facilities, Safety Series No. 17, IAEA, Vienna (1966).

[11 ] BARRY, P .J ., "Concept o f a standard site", Containment and Siting o f Nuclear tower Plants (Proc. Conf. Vienna, 1967), IAEA, Vienna (1967) 205.

[12] INTERNATIONAL ATOMIC ENERGY AGENCY, Economics in Managing Radioactive Wastes, Technical Reports Series No. 83, IAEA, Vienna (1968).

56


APPENDIX 1


APPENDIX 1

RADIOACTIVE WASTE MANAGEMENT AT CANADIAN NUCLEAR POWER REACTORS

C.A. MAWSON Biology and Health Physics Division,

Atomic Energy of Canada Limited,Chalk River, Ontario, Canada

PART I

INTRODUCTION AND GENERAL DISCUSSION

This paper contains Part I, which is a general introduction, and Part II, which consists of reports written by the staff of the two power reactors at present operating in Canada.

The reports in Part II are each arranged according to the scheme suggested by the IAEA. The numbering, where applicable, accords with this scheme.

NPD is a 25MW(e) demonstration station which has been delivering power to the grid since 11 April 1962, so considerable experience has been accumulated. Douglas Point (200MW(e)) is still in the commissioning stage but started delivering power regularly on 7 January 1968 at 150MW(e), which was raised to full power on 8 March. During the earlier stages of commissioning the station operated at less than full power for varying periods, beginning in December 1966.

The power reactors in operation at present are of the heavy- water, natural uranium type. The fuel is cooled by heavy water passing through pressure tubes, with transfer of heat to light water for driving the turbines. The spent fuel, which is removed from the reactor on power, is stored under water and is not reprocessed. A 2000MW(e) station under construction at Pickering, near Toronto, is also of this type. The station being built at Gentilly, between Montreal and Quebec, is also a heavy-water, natural uranium r e actor but the coolant is boiling light water, which drives the turbine.

59


Principles of Waste Management

The object of waste management is to dispose of waste in such a way that the resulting dose of radiation to members of a "critical group", as defined by the International Commission on Radiological Protection (ICRP Publ.7, 1965), will be less than the maximum permitted by the recommendations of the Commission (ICRP Publ.9, 1965).

Each station is studied as a special case, and a realistic assessment is made of the dilution capacity of the environment, the location and relevant habits of the population and the quantity and nature of wastes likely to arise. The waste-management system is designed to deal safely with these wastes in the environment that actually exists or can reasonably be forecast to exist during the lifetime of the reactor, taking into account the kinds of accidents that might be expected to occur. Provision is not made for handling wastes arising from a "Maximum Credible Accident", although the consequences of such a hypothetical event are considered in drawing up the Emergency Plan. This forms a part of the Safety Report submitted to the Atomic Energy Control Board during the licensing procedure.

In some cases the nature of the ground, the distribution of population and the possibilities of contamination being transmitted to man have made it practicable to locate at the site a ground burial area for reception of normal solid wastes. At other sites the transportation of such wastes to an established disposal area is practised or intended. At two of the larger power stations, spent ion-exchange resins are stored in underground tanks designed to be adequate for the lifetime of the reactor. At the Gentilly reactor the ion-exchange resin from the boiling-light-water circuit will be more bulky than at the heavy-water-cooled plants, and it will not have to be dried, so it will be incorporated into concrete blocks. In all cases, high- level solid wastes unsuitable for local disposal are, or will be, transported to ah established waste disposal area.

All Canadian power reactor sites are on large bodies of water - the Ottawa River, the St. Lawrence River, Lake Huron and Lake Ontario. The permitted release rate of radionuclides is, or will be, in accordance with the results of environmental studies leading to knowledge of the dilution capacity of the environment, thus ensuring that the provisions relating to maximum perm issible dose to critical groups are fulfilled. At present the concentration of

60


fission products derived from weapons fall-out is large in these waters compared with the added contamination from power plant effluent. The discharges into the atmosphere are negligible as a hazard to man, in fact the two stations under construction do not have tall stacks. The major concern has been to keep the gaseous effluent away from the ventilation intakes..

The waste-management systems are essentially simple. The system for liquid wastes consists of receiving tanks followed by dispersion tanks arranged in such a way that a tank in use is matched by another tank which is full. The latter tank is sampled and an analysis is completed before the contents are passed on. After re lease from a dispersion tank the effluent has to pass a monitor which, if the radionuclide content is above a set level, returns the effluent to the tank. After passing the monitor the effluent enters the condenser cooling water. This is sampled by a proportional sampler before discharge to the river or lake.

Segregation of wastes into "high-level" and "low-level" streams ensures that if a receiving tank contained exceptionally high-level waste it would be directed to a concreting plant where it would be mixed with cement in steel drums and dealt with as solid waste. This system has not, in fact, been called upon to operate, as such wastes have never arisen. It has always been possible to discharge the "high-level" wastes through the dispersal tank system. The valving of the whole plant is arranged so that waste can be directed from any place in the system to any other place.

The active gaseous waste is filtered through absolute filters before discharge to the atmosphere. The filters are dealt with as solid waste. Provision is made for decay storage of high-level gaseous waste, but this has never been necessary. The radioactive components of the stack effluents are almost entirely tritium and 41A. Up to now we have had only four defective fuel bundles in our power reactors, all due to mechanical damage. Because of on- power fuelling, these caused no operating problems and some were left in the reactor until their turn came for discharge. There was no serious contamination of the coolant.

Our experience has been that the amount of radioactive waste produced by a nuclear-power reactor is so small that no "problem" exists. It can be dealt with by well-known techniques, mainly those dependent upon controlled release to the environment. Recognizing the fact that the potential for such releases will decrease with the multiplication of sources of radioactive contamination, we do not foresee serious restriction from this cause in the near future.

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PART II

REVIEW OF WASTE MANAGEMENT AT THE NPD GENERATING STATION

I. INTRODUCTION AND GENERAL DISCUSSION

1.1. Basis of waste-s.ystem design

Solids

As waste disposal of solids is by shipment to Chalk River National Laboratory (CRNL), no problems are expected in regard to the capacity of the system.

Liquids

The installed system is adequate.

Gases

Improvements in monitoring are planned, otherwise the in - stalled system is adequate.

.The waste disposal methods are shown in F ig .l .

1.2. Limitations

Solids

No radioactive solids are disposed of on the plant property; all are shipped to CRNL. Limits exist only on the intensity of radiation from these shipments.

Liquids

The maximum permissible concentration of 90Sr in effluent entering the Ottawa River is 1 X 1 0 '6/LjCi/ml.

REPORTS FROM OPERATING STATIONS

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Gases

(1) Maximum permissible concentrations in air (averaged over one year) at ground level at a point outside the controlled area:

Maximum permissible Maximum permissible concentration release (Primary Standard)

(Ci/m3) (Ci/d)

Noble gases (41A)131 j

Tritium (as oxide)90Sr 131Cs

3 X io -8 5 X lO"13 2 X lO '81 x n r 112 X 10 -11

3 X 103 0.1

4 X 103 1.04

The maximum permissible concentrations in column 2 will not be exceeded (annual average) if the average daily releases shown in column 3 are adhered to.

(2) The following maximum permissible doses are allowed at any one time in addition to those resulting from the maximum permissible concentrations above providing they do not occur more than four times a year.

Nuclide

Noble gases 13 l j

Tritium (as oxide) 9°S r

137Cs

Maximumpermissible

dose(Ci-s/m 3)

3 X lO’ 11.3 X 10’ 6

1 X 10-11 X 10-54 X 10‘ 5

Maximumpermissible

release(Ci)

3X 1042.5 X 10-1

2 X 1041 8

1.3. Capital and operating costs

Capital cost for handling spent fuel (solids) was 0.2% of plant cost; operation was 0.7% of plant operation cost in 1967.

Capital cost for other solid waste handling equipment was 0.1% of plant cost; operation was 0.8% of plant operation cost.

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05

L t Q U t O S

I K ? LMDS

FIG. 1. NPD active waste disposal methods.


Capital cost for liquid waste handling equipment was 0.05% of plant cost; operation was 0.06% of plant operation cost.

Capital cost for gaseous waste handling equipment was 0.3% of plant cost; operation was 0.7% of plant operation cost.

The costs are high because NPD is an experimental and demonstration station, and thus produces more wastes than would be expected from a normal nuclear power station.

1.4. Operational experience

Solids

The manpower requirement is approximately 830 man hours per year at NPD. Maintenance of facilities is slight. There have been no unusual incidents. The handling rate for low-activity solid wastes doubles during reactor shutdown periods due to increased work on modifications and maintenance of process systems. The only desirable design change would be an increase of shielded fa cilities for temporary storage.

Liquids

The manpower requirement is approximately 360 man hours per year. Maintenance consists of routine pump service and sump cleaning. There have been no unusual incidents. There are no significant changes in the operation of this system due to reactor operation. The installed equipment was changed to a manually controlled batch disposal system because the low volumes and low activities of liquid waste did not justify perfection of the automatically controlled system originally installed.

Gases

The manpower requirement is approximately 1500 man hours per year. Maintenance consists largely of fan servicing and absolute filter replacement. There have been no unusual'incidents. Reactor operations result in a release of 41A. It is intended to im prove the monitoring systems for the stack effluent.

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11.1. Category 1 ( < 200 mR/h on contact)

Sources and quantities

This includes radioactive scrap metal, rags, glass, sumpdregs, wood, paper and used air filters. Approximately 25 000 kg were handled in 1967.

Collection, treatment and disposal

This material is generally collected in paper bags which are subsequently stored and shipped to CRNL in steel drums.

11.2. Category 2 (200 mR/h to 1 R/h on contact)


This includes scrap and crud from the fuelling machines and process systems. Approximately 140 kg of Category 2 waste were handled in 1967.

Collection, treatment and disposal

This material is handled in the same manner as Category 1 material.

11.3. Category 3 ( > 1 R/h on contact)


In 1967 this included disposal of 1200 kg of ion-exchange columns, 1000 kg of reactor components (almost entirely due to the experimental replacement of two pressure tubes) and 1000 kg of spent fuel assemblies.

Collection, treatment and disposal '

In general the material is collected in shielded containers for shipping. Spent fuel is generally stored for a one-half year decay period before shipment to CRNL.

II. SOLID WASTES (in three categories)

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III. LIQUID WASTES

111.1. Sources and quantities

There are organic liquid wastes from decontamination work and the control laboratory. This waste volume was 280 litres in 1967. Aqueous radioactive waste sources include the laundry (7000 litres/yr); the spent fuel bay (175 000 litres/yr); low concentration heavy water (70 litres/yr); subsurface drainage (70 litres/yr); and other drainage from decontamination and the control laboratory. The total release of low activity water to the Ottawa River was approximately 1 900 000 litres in 1967.

111.2. Collection, treatment and disposal

The organic liquid is collected in drums and shipped to CRNL. Aqueous radioactive waste drains, oris pumped, to a4500-litre sump in which settling takes place. The sump is pumped to a 45 000-litre tank. The contents are sampled, then released at controlled rates into the normal plant process drainage system which provides adequate dilution. A provision exists for pumping the contents of the tank through an external connection to a tank truck or pipeline, but this facility has never been required. Construction is generally of cast iron and steel. Pumps have bronze impellers and range in size from 1/2 hp to 5 hp (3.8 litres/s). The activity release by this system in 1967 was: tritium, 3700 Ci; gross beta activity, 180 mCi; 89Sr plus 90Sr, 24 mCi; 144Ce, 6 mCi; 106Ru, 3.7 mCi; W Cs, 12 mCi.

The sump dregs represent approximately 0.1 C i/yr and are handled as Category 1 solid w aste.-

IV. GASEOUS WASTES

IV. 1. Sources and quantities

All exhaust air from rooms with potential for activity release is filtered, then discharged via the 50-m stack. The maintenance of favourable ventilation pressure differentials necessitates exhaust of air from the reactor vault and the boiler room. Consequently there are releases of 41A and tritium. Fission products are a potential source; however, 131I release has not exceeded 1700 p Ci/yr. The normal stack exhaust flow is 25 000 ft3/min.

67


IV .2. Collection, treatment and discharge

An off-gas delay tank exists but has not been required. Thirty- six absolute filters are installed. Adsorption filters exist only for emergency operation of the containment system. Exhaust ductwork is of all-welded construction. Fans are fitted with mechanical shaft seals.

Releases in 1967 were higher than usual. The tritium release was 9000 Ci. The 41A release was 3400 Ci.

REVIEW OF DOUGLAS POINT GENERATING STATION WASTE MANAGEMENT

I. GENERAL

1.1. Design basis

The design of the solid-waste system is based on normal operation for 30 years, with capability of extension. The liquid- and gaseous-waste systems are sized for maximum expected requirements. The liquid-waste disposal system is shown in F ig.2.

1.2. Limitations

1.2.1. Permissible release levels in air

Maximum permissible releases (averaged over one year) are given in Table I.

The maximum permissible instantaneous releases permitted in a year, in addition to the above, are shown in Table II.

For combined release the permissible level is given by

Ci 137Cs , Ci 3H , Ci N.G. „ ,5 7 X 103 5 X 10 -

1.2.2. Permissible release levels in liquid effluents

The maximum permissible releases (averaged over one year) are given in the Table III.

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NORMALLYINACTIVE

RINSES & FLOOR

CHEMCALLY ACTIVE

DEC0N7AM & LAB BAY

FIG. 2. Douglas Point liquid waste disposal system.

TABLE I. MAXIMUM PERMISSIBLE RELEASES IN AIR (averaged over one year)

RadionuclideMaxim um perm issible release

(C i /d ) a

lod in e-131 4 X 10 '2

Strontium-90 6 X 1 0 '1

Caesium -137 5

Tritium (ox ide) 7 X 10s

N oble gases 5 X 10s

a The instantaneous release rate, under normal conditions, shall not exceed ten tim es the average release rate in C i /s ca lculated from the above tab le.

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TABLE II. MAXIMUM PERMISSIBLE INSTANTANEOUS RELEASES IN AIR

RadionuclideIntegrated annual release

(C i)

Iodine-131 8 X 1 0 '2

Strontium-90 1

Caesium -137 10

Tritium (oxid e) 1 .5 X 104

N oble gases 1 X 1 0 4

TABLE III. MAXIMUM PERMISSIBLE RELEASE IN LIQUID EFFLUENTS

RadionuclideRelease(m C i/d )

Maxim um concentration in discharge ( f jC i /c m 3)

Iodine-131 3 X 1 0 2 1 .5 X 1 0 '6

Strontium-90 4 X 1 0 2 2 X 1 0 " 6

Caesium -137 1 X 1 0 2 5 X 1 0 " 7

Tritium (ox id e ) 3 X 106 1 .5 X 1 0 " 2

C obalt-60 8 X 1 0 3 4 X I O '5

Gross beta 10 3 5 X I O '6

For combined release the permissible level is given by

mCi 137Cs mCi 3H mCi 60Co _10 0 3 X 10 6 8 X 103 -

1.3. Capital and operational costs

The capital cost of all waste-management systems was 0.408% of the total station cost, and the operating cost, measured through

70


the year 1967, was 0.421% of the total station operating cost. This, however, includes commissioning operations on waste-management systems.

I.4. Operational experience

The operating experience has so far been limited, since the station is still in the commissioning phase. However, all operational and maintenance manpower requirements for managing wastes have been provided by station operating personnel incidental toother operations, without specialized assistance. There have been no unusual incidents in disposal operations. A change in management of solid waste was made by the introduction of burning low -level waste contaminated with activation products in an open pit at the waste disposal area. Another change is contemplated to introduce heavy-water recovery equipment in the ventilation exhaust system that will reduce tritium releases. These changes are incorporated in the information below.

II. SOLID WASTES

11.1. Sources, types and quantities

The general categories are: ion-exchange resin; system air and ion-exchange filters; contaminated and/or activated equipment; and miscellaneous contaminated plastic, rags or paper waste (e.g. cleaning wastes). Additionally, concreted liquid waste Is handled as a category of solid waste. Irradiated fuel is not disposed of at the station but stored under water.

Annual quantities of waste cannot yet be established as waste disposal operations only began in June 1967. Quantities since that time are given in Table IV.

11.2. Method of collection

Miscellaneous waste is segregated at source into specially identified collection cans lined with yellow 4 -m il plastic disposal bags. The bags are replaced routinely and stored at one collection point. Equipment is wrapped and stored at the same location im mediately it is designated as waste. Ion exchange resin is fed directly into a disposal pit under the reactor building.

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TABLE IV. QUANTITIES OF W ASTE

Filters & equipment M iscellaneous contam inated waste

Month N o. o f item s A ccum ulative activity (C i)

V olum e( ft 3) a

A ccum ulative activity (C i)

Jun. 67 8 0.2740

Jul. 67 5 0.2850

A ug. 67

Sep. 67 2 0.2888

O ct. 67 12 0.3433 304 0.6525

N ov. «7 3 0.5050 108 0.7241

D e c . 67 162 0.75973

Jan. 68 22 0 .6331 122 0.79153

Feb. 68 17 0 .9039 122 0.85763

Totals (480 ft3) 0 .9039 818 0.85763

N ote : A ctiv ity in the above is activation products. A ctiv ity for bags o f waste is ca lcu la ted on the experim ental basis o f a bag o f contam inated ra g s (0 .5 m Ci o f 137Cs) measured for average gamma dose-rate at 1 ft from the surface and at contact, representing 1 m Ci as 1 .2 m R/h, at 1 ft. A ctivated equipment is measured on the basis o f 1 m Ci representing 4 .0 5 m R/h at 1 ft, be ing the dose-rate derived from sealed 137Cs sources.

a 1 fts= 0 .028 m 3 b 1 ft = 3 0 .5 cm .

II.3. Treatment

Bags of miscellaneous waste that are too active for burning (see Section II.4) are baled in a hand-operated baler. Concreted waste is dealt with under the treatment of liquid waste.

II .4. Disposal

Waste is disposed of into one of four facilities at the station waste disposal area, more according to type than activity categories.

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The disposal area comprises about 1.5 acres, about 1 mile inland from the station. The facilities and types of waste intended for the area are listed below:

Open burning pit For combustible wastes with activation- product contamination up to 25 mR/h at 1 ft from container bags

Concrete trenches For baled waste (exceeding 25 mR/h at 1 ft), contaminated and/or activated equipment and filters, up to 50 R/h at contact

Lined holes For filters and equipment exceeding 50 R/h at contact

Monoliths For steel drums of concreted liquid waste

III. LIQUID WASTES

III. 1» Sources, types and quantities

Liquid waste categories and design quantities are listed below.

Category Source

Normally inactive Laundry and washrooms

Active Decontamination rinses and building floor drains

Chemically active Laboratory anddecontamination solutions; spent fuel storage bay slurry

Imp, gal Vmonth

42 000

27 000

630

1 1 Imperial gallon = 4.55 litres.

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III.2. Collection and treatment

"Normally inactive" waste is admitted to the Service Water effluent after passing an in-line activity monitor. The flow is automatically redirected to a hold-up tank if activity exceeds the monitor set point.

"Active" wastes flow directly to the hold-up tank and are held for decay and analysis. Low-activity liquid goes to one of two dispersal tanks from which it is fed to the Service Water effluent at an acceptable activity rate. High-activity liquid is diverted to a feed tank.

Chemically active waste is fed directly to the feedtank. The contents of this tank are neutralized if necessary, then disposed of by dilution and release or, if this is not possible, they are processed.

Processing is by remote-controlled concreting in 45-gal steel drums, and when hardened, handled as solid waste.

If volumes demand, provision is made for the installation of an evaporator.

All initial collection lines and inaccessible headers are made of copper. Accessible pipes are carbon steel. All tanks are steel excepting the feed tank, which is stainless steel. The two dispersal tanks are each 40 000 Imp. gal, the hold-up tank 7000 Imp. gal, and the feed tank 2000 Imp. gal.

III.3. Discharge to environment

The discharges by volumes and activity concentration at the outfall are shown below for three months representative of operation at approximately 75 to 80% of full reactor power. The activity concentrations are the average gross beta measurements from weekly samples taken from the outfall proportional sampler.

Month

Average gross beta QuCi/litre)

Effluent volume (Imp, gal)

Dec. 1967 Jan. 1968 F e b .1968

4.2 X 10"6 31.6 X 10-6

8.5 X 10-6

9480 X 106 7938 X 106 7900 X 106

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IV. 1. Sources, types and quantities

IV . GASEOUS WASTES

Gaseous wastes are from the following sources:

Reactor building

Thermal shield cooling system bleed; excess volume and purges from closed systems (vaults); primary heat transport system off- gases holding tank; air from accessible work areas.

Service building

Flow from laboratory and decontamination centre fume-hoods; air from waste management and spent fuel storage rooms; air from active workshops.

The volume from the reactor building is 8 X 106 cm3/s andfrom the service building 20.3 X 106 cma/s .

IV.2. Treatment

The reactor building ventilation exhaust acts as a collector for all sources: the system does not provide holding for decay except for the primary heat transport off-gases. A triplicated monitor on this exhaust (reactor building exhaust) will shut down the fans and close the ducts when two out of three alarm. A by-pass allows building overpressure to be released to the service building exhaust system to prevent leakage from the reactor building. This by-pass is provided with a demister and iodine filter. The main building exhaust passes through an automatic roll-type prefilter and a bank of high-efficiency filters: (99.97% on particles of 0.3 Mm)-

The service building exhaust flow from fume-hoods with absolute filters is fed directly to exhaust fans, and the remainder is fed through a prefilter and absolute filter sim ilar to the reactor building exhausts.

The exhaust from both buildings is then combined and released to the stack. On the combined duct there are two sample flows: one through an iodine monitor, and the other through a particulate monitor, then a noble-gas monitor. All monitors have control room annunciation. Tritium is also measured from a heavy-water gas washer sampler.

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IV .3. D ischarge to environment

To date, no radioiodine and no particulate activity have been detected in the gaseous effluent. The noble-gas monitor has not yet been equipped with an integrator, necessary for measuring a ccumulative releases. However, in April 1967 a measurement of argon activity in the source atmosphere and the monthly volume re leased from it showed a typical monthly release value of 350 Ci.In the past, tritium has been released to the extent of 203.4 Ciduring 1967. The values for more recent months are:

Dec. 67 - 96.7 CiJan. 68 - 88.0 CiFeb. 68 - 102.0 Ci

Planned conversions to a dry vault system for the vault ventilation system and the installation of a dryer on the boiler room exhaust will end this release rate in the near future.

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APPENDIX 2


AP P E N D IX 2

RADIOACTIVE WASTE MANAGEMENT AT NUCLEAR POWER PLANTS IN FRANCE

G. COHENDY

CEA, Centre de Production de Plutonium de Marcoule, France

In France, the nuclear power reactors are essentially of the uranium-fuelled, graphite-moderated, gas-cooled type. Two of them, of the G-series (G for graphite), are located at Marcoule (Gard) and another three, of the EdF series (EdF = Electricite de France) are in operation at Chinon (Indre et Loire). However, the French program is by no means limited to this class of reactors, and a uranium- fuelled, heavy-water-moderated, gas-cooled reactor is in operation at Brennilis (Finistere), and a PWR at Chooz (Ardennes) as a joint Franco-Belgian venture. Furthermore, an experimental sodium- cooled breeder, theRAPSODIE, has been in operation for more than a year at Cadarache (Bouches-du-Rhone).

Although the French reactors are of various types, the waste- management problems are more or less the same at all sites and the differences are of minor importance.

I. INTRODUCTION AND GENERAL REMARKS

I. 1. Principles of waste management

All French nuclear power plants are equipped for solid- and liquid-waste treatment, but the degrees of development and utilization of the existing facilities vary from one case to another. In some centres (at Chooz and Brennilis), the reactors have not been in operation long enough to give rise to specific problems, in others the facilities are fairly sizeable (Chinon, for example). The G2-G3 plants and RAPSODIE are located on multipurpose sites; Marcoule is a plutonium production centre and Cadarache is devoted to research and development work in various fields. For this reason, the size

79


of the waste-treatment facilities at these two sites is out of proportion to the importance of their reactors.

Generally speaking, all French nuclear plants are properly equipped for radioactive waste management. Thus, all sites possess the storage facilities, especially for liquid effluents, needed to cope with emergency situations. Solid and gaseous wastes are often stored to allow radioactive decay before treatment or disposal.

1.2. Limitations

Some limitations, essentially of a legal nature, implying observance of certain rules by the operators, are imposed on waste management.

Thus, gaseous wastes are released to the atmosphere via specially designed stacks and diluted to a level always below the maximum permissible concentration for the public.

The characteristics of packaging for the transport of radioactive substances have been fixed by law. Solid radioactive wastes also are subject to these regulations, one of the main provisions of which is that the dose-rate shall not exceed 200 mR/h at the surface of the package (and 10 mR/h at a distance of 1 m fpom it). To preclude the need for providing additional protection during any transportation outside the sites, plant managements have almost without exception packed their solid wastes in such a manner that dose-rates do not exceed the limits indicated above.

All liquid-waste disposal is under the control of the Ministry of Public Health, Central Service for Protection against Ionizing Radiation (S. C. P. R. I. ). All waste disposal must be in accordance with the general formula:

L ------------ — ------------ ^ 1(C M A P )j

where CMAP = maximum permissible concentration for the public.Thus, there is a legal limit, in addition to a geographical limit,

when the particular disposal formula of a plant takes into account the flow of the river into which discharge takes place, such as that of the Rhone at Marcoule. It should be mentioned that the discharged activity everywhere is distinctly below the imposed limit. Even at

8 0


Marcoule, where it depends only slightly on the liquid wastes from ■ reactors, it has never exceeded a few per cent of the activity authorized under the disposal formula.

I. 3. Capital and operating costs

These are not well known, either because the experience of the power plants is not adequate to provide a sound answer, or because most of the wastes treated are not from power reactors.

Nevertheless, an order of magnitude can be indicated for the G-series reactors. The investment costs for waste processing plants are 1% of the investment in the reactors themselves; the proportion for operating costs (including amortization) is somewhat higher, but is of the same order of magnitude.

I. 4. Operating experience

This is fairly limited for experimental reactors (uranium-heavy- water-gas, pressurized-water and breeders).

On the other hand, considerable experience has been built up with the uranium-graphite-gas-type reactors. Whether the reactors are part of a diversified plant, as at Marcoule, or are the very reason for the existence of the plant, as at Chinon, operating experience with them has shown that the facilities were well designed and that they are adequate even in the event of an incident. Therefore, no modifications are necessary in the original system.

II. SOLID WASTES

II. 1. Source, type and annual amounts

Solid wastes originate either from decontamination operations or from dismantling in connection with changes made in the reactor.

II. 1. 1. Wastes produced by decontamination operations

During action following an incident, it is necessary to install temporary floor and wall coverings, which usually consist of sheets

8 1


' of polyvinyl chloride (PVC) (or a more readily burnable material when the Centre possesses an incinerator). Decontamination work proper produces contaminated wastes, such as gloves, cotton, rags, etc.

II. 1.2. Wastes from dismantling and other operations

Another type of waste consists of contaminated and, in particular, activated pieces of metal. Serious problems are frequently caused by the induced radioactivity following the removal or replacement of pieces which have been in an appreciably high neutron flux. Another category of wastes consists of contaminated rubble or earth resulting from the rupture of an active effluent channel, for example.

II. 1.3. Amounts produced

No experience has been acquired with other systems, but some precise figures are available for uranium-graphite-gas-type reactors. Thus, in the case of the EdF reactors, about 200 m3 of wastes were produced in 1967 for an installed power of 700 MW(e). These wastes were stored in 200-litre drums, whose surface dose-rate is less than 500 mR/h.

In the case of the G -series reactors, again in 1967, roughly 63 m3 of compressible and 75 m3 of incompressible wastes were produced.

II. 2. Method of collection

II. 2. 1. Transport containers

At all the plants, wastes are collected at the actual source, either in disposable drums, or in recycled stainless-steel bins, with a capacity of 100 or 200 litres. When the dimensions or type of the wastes is such that it cannot be placed in the collecting drums or bins, it is packed in double or triple thicknesses of a suitable material (e. g. PVC) and usually disposed of in a leak-proof tank or suitably protected container, if its activity is too high for normal handling procedures.

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II. 2. 2. Manner of collection

Collection is systematic, i. e. it takes place on fixed days or, if necessary, as soon as the volume accumulated justifies it. Wastes are usually sorted into categories (combustible, incombustible but compressible, incombustible and incompressible) at the actual sources.

II. 2.3. Means of transport

At all sites, the wastes are removed by truck, either to the waste-treatment plant, or to the storage facilities.

II. 3. Treatment

The purpose of the treatment is maximum volume reduction of the wastes for containment.

II. 3. 1. Compaction

Considerable reduction in volume can be obtained by compaction and it is used at most of the plants; compaction pressures are usually about 35 kg/cm2. The wastes are then incorporated into concrete, either in 200-litre drums of the type used for oil (which results in units that are easier to handle, since they do not weigh more than 500 kg) or in prefabricated concrete containers with an internal capacity of 700 litres and a weight of 2. 5 t.

II. 3.2. Incineration

This method affords the best volume reduction factors. However, the necessary installations are more complicated than those needed for compaction; in particular, it is absolutely essential to have a waste-sorting installation to ensure as far as possible that only combustible wastes are fed into the incinerator.

There are incinerators at two of the centres, Marcoule and Cadarache. The Marcoule unit can handle 80 kg/h of solid wastes and 30 litres/h of oil simultaneously. The waste is burned as a thin

83


layer and the gases pass into a post-combustion chamber before entering an air-cooled heat exchanger. Off-gas purification is by the dry process and the incinerator is batch-fed.

The Cadarache incinerator has an output of 30 kg/h. Like the Marcoule plant, it has a post-combustion chamber and is batch-fed. However, gas purification is by the wet process. For the moment, this installation cannot be used for contaminated oils.

II. 3. 3. Fragmentation

Certain incombustible wastes are not suitable for compression, particularly because of their size or mechanical features, and therefore have to be broken up so that they can be included in the normal processing cycle. This applies particularly to metal pieces, which are cut either with alligator shears, a frame-saw or a pipe-cutter. They can also be broken up with a blow torch, either in the open air, or in a pool, as at Marcoule, where a plasma blowpipe is used; this process gives excellent results and provides maximum possible protection against irradiation hazards and dispersal of contamination.

II. 4. Disposal

The problem of disposal has been solved in a number of ways. Generally speaking, a distinction can be made between EdF-type reactors (as well as one reactor of the uranium-heavy-water-gas- type) and G-series reactors (and one pressurized-water reactor and breeder).

II. 4.1. EdF-type reactors

The wastes from these reactors undergo no processing; they are stored either in drums or in appropriate packagings. The most active wastes are compacted at a pressure of 45 kg/cm 2 and are stored inside wells from which it is assumed that they will not be removed. For example, the total volume of the wells at Chinon is about 1500 m3 , of which only 16 m3 are at present used. Low-activity wastes may be removed later for final storage. Their present volume at Chinon is less than 500 m3. Before storage, they are

84


merely subjected to a summary reduction in volume by means of a crusher.

II. 4. 2. G -series reactors

As already stated, low-activity wastes are incorporated in concrete after compaction or fragmentation (and even incineration, as far as ash is concerned). During 10 years' operation at Marcoule, about 2500 m3 of low-activity reactor wastes have been treated in this manner. The active wastes are stored in holding trenches until they can be introduced into the regular processing cycle.

II. 4. 3. Disposal outside the sites

No decision has been made as yet in France regarding the final disposal of these wastes, which are at present stored on the sites.

III. LIQUID WASTES

III. 1. Source, type and annual amounts

The sources of liquid wastes vary considerably, depending on the type of reactor.

III. 1. 1. Uranium-graphite-gas reactors

These reactors themselves produce no effluents of any kind (except in very special cases, such as the EdF-1 thermal shield). The only effluents are "suspect" ones, such as those resulting from soil leaching in the active zone, from laundries, equipment, decontamination shops, etc. The volume of these effluents is about 20 000 m3 annually at Marcoule and 100 000 m3 annually at Chinon.

III. 1.2. Uranium-heavy-water-gas reactor

Heavy water raises a special problem and all effluents likely to contain it are subjected to expensive recovery operations; this

8 5


water also contains appreciable amounts of tritium. Another special category of effluent is that consisting of residues from the regeneration of ion-exchange resins.

III. 1.3. Pressurized-water reactor

The water of the primary circuit contains boric acid which provides a chemical control system for the reactor. This water gives rise to a number of problems,especially because of its expansion during reaction start-up. It is also hydrogenated to minimize corrosion by reducing the oxygen concentration resulting from radiolysis; this raises a further problem for waste management.

III. 1.4. Breeder reactor

The only problem here is that caused by the wastes from decontamination operations; no water is permitted within the reactor containment. Moreover, the effluents are in very small amounts, as the main decontamination agent is alcohol, which is distilled and re-used.

III. 1. 5. Contaminated oils and solvents

These liquid wastes cause a special problem which can be reliably solved only by incineration or distillation when homogeneous solvents are involved. However, it must be pointed out that the amounts produced are small, e. g. about 5 m3 of oil annually for the three EdF reactors at Chinon.

III. 2. Collection and treatment

III. 2. 1. Characteristics

All liquid wastes from reactors are characterized by a low specific radioactivity, which does not exceed 10-5 Ci/m3 . The only exception is the water of the primary circuit of the PWR, for which the value is of the order of 10"2 Ci/m3 .

86


III. 2 .2 . Collection and storage

The two sites (Marcoule and Chinon) at which the uranium- graphite-gas reactors are located are equipped with a pipeline system; elsewhere, effluents are collected by tanker trucks.

There are large storage areas at all the plants; e.g . more than 2000 m3 capacity for the three EdF plants at Chinon and over 1000 m3 for the pressurized-water plant at Chooz. Generally speaking, these facilities are designed mainly for convenience in operating. The degree of deactivation is such that the problem of degassing does not arise even at the pressurized-water plant, where the vents are connected to the gaseous effluent storage facilities.

III. 2.3. Treatment

Except at the pressurized-water plant, where primary-circuit water is treated by an evaporator and then by an ion exchanger, the nature of the effluents is such that disposal into the environment calls merely for simple physical filtration. The evaporator for treating the prim ary-circuit water at Chooz has a working capacity of1.5 m3 and can process 3.5 m3 hourly; it is made of stainless steel. There is a smaller evaporator at Chinon (1 m 3/h), but this has not yet gone into operation. A third, very similar evaporator is used at Brennilis.

III. 3. Disposal into the environment

After processing, liquid wastes are discharged into the environment. Except in the case of the Ellez, at Brennilis, which is only an average-sized river, there are no problems arising from flow, even during low-water periods, because the rivers involved are all large: the Rhone at Marcoule, the Loire at Chinon and the Meuse at Chooz; even at Cadarache, the flow of the Durance is sufficient to prevent disposal problems.

The activity released is always very low, e .g . about 1 C i/yr at Chinon for 100 000 m3 of effluent (and an installed power of 700 MW(e)). However, it is higher at Chooz, because of the presence of activating isotopes (especially ^Co, 59Fe and 60Co, and also 3H).

87


III. 4. Waste disposal

The concentrates from the treatment of liquid wastes enter into the solid-waste processing or storage cycles. At Chooz, for example, they are incorporated in concrete and placed in the same type of prefabricated concrete containers used for solid-waste disposal. At Brennilis, evaporation wastes are stored in crim ped drums. At Chinon, it is planned to store sludges in three 6. 5 m3 drums of unit capacity once the evaporator is put into operation. When one of these drums is full, it will be placed in a holding well so that, a priori, the total activity — and thus the dose-rate — at the surface of the drums is not , limited. In all other cases, the criterion applied is the same as for solid wastes, i. e. the dose-rate must not exceed 200 mR/h at the surface of the storage container and 10 mR/h at a distance of 1 m.

IV. GASEOUS WASTES

IV. 1. Source, type and annual amounts

The ventilation air from the installations, discharged into the atmosphere after filtration, is mentioned merely for the record.

The problems connected with gaseous wastes vary from one type of reactor to another.

IV. 1.1. Uranium-graphite-gas reactors

The main difficulty in these reactors is leakage of carbon dioxide or the periodic emptying of sections of the cooling system. The latter operation is handled in two ways. Coolant from EdF-type reactors is discharged directly into the atmosphere through a stack above each reactor; in the case of G -series reactors, the carbon dioxide is completely liquefied and stored for re-use.

IV. 1. 2. Uranium-heavy-water-gas reactor

Here again with these reactors, the carbon dioxide problem arises (the gas is discharged into the atmosphere when necessary). However, this is less important than the problem caused by the cool-

88


ing air passing close to the core with the result that its argon is appreciably activated (3 Ci/MW per d). In the event of a pressure tube fracture, there would be a release of tritiated heavy-water vapour.

IV. 1.3. Pressurized water reactor

The gases In this type of reactor emanate from the primary- circuit vents; as already mentioned, this means that there are hydrogenated gaseous wastes. These are stored in a nitrogen atmosphere and their oxygen content is monitored.

IV. 1. 4. Breeder reactor

The blanket argon in these reactors obviously contains ^A, and also 133Xe, 135Xe and 23Ne, from an (n, p) reaction with sodium. On the other hand, there are practically no activation products in the nitrogen used to cool the shielding.

Except in the case of the PWR, where the vents from the liquid wastes are connected to the gaseous-effluent storage facility, there are no problems from gases from liquid wastes. Nor do the off- gases from incinerators cause any difficulties, thanks to the very thorough filtration. At Marcoule, for example, disposal is through a stack only about 20 cm higher than the roof; the activity of the discharged gases is of the order of 10"12 C i/m 3, and the maximum activity has never exceeded 10-11 C i/m 3 .

IV. 2. Collection and treatment

IV. 2.1. Uranium-graphite-gas reactors

No collection or treatment is necessary in the case of EdF-type reactors, since the carbon dioxide is discharged directly into the atmosphere. In the case of the G -series reactors, this gas is collected, liquefied and re-used after simple filtration, withoutspecial treatment.

8 9


IV. 2 .2 . U ranium -heavy-w ater-gas reactor

This reactor has activated charcoal filters in all circuits likely to transport iodine.

IV. 2.3. Pressurized-water reactor

The gaseous effluents in this reactor can be stored for de-activation in four 44 m3 tanks which can withstand a pressure of 10 b.

IV. 2 .4. Breeder reactor

The radioactive gases in this reactor can also be stored for deactivation in two 6 m3 balloons, and it is planned to trap xenon on active carbon before discharge to the stack.

IV. 3. Disposal into the atmosphere

This always takes place through stacks, the most original and spectacular of which is that of Chooz, which is cut through the rock in the very centre of a hill overlooking the Meuse and has an internal diameter of 2. 25 m. It is more than 200-m high. About 2000 mP of waste were discharged in 1967, the activity after de-activation having been about 2 Ci.

The RAPSODIE stack at Cadarache is 45-m high and discharge takes place through absolute and iodine-retaining filters.

The stack at Brennilis is 70-m high and can discharge all the off-gases from the leak-proof reactor containment, the entire coolant of the reactor (corresponding to 14 t of carbon dioxide) and all escaping gases (about 2 t daily); as already mentioned, the main gaseous radioisotope is 41A.

Carbon dioxide leakage at Marcoule corresponds to a monthly activity of 450 Ci of 41A for each reactor. There is no other discharge into the atmosphere.

Generally speaking, it is difficult to estimate the amounts and activities discharged. In most cases, there is uncontrolled leakage or else the gaseous effluents are mixed with the off-gases in general.

90


Only at Chooz is it possible to obtain an accurate idea of the amounts involved, since the radioactive gases are stored before discharge.

V. CONCLUSION

This review of the problem of radioactive wastes produced by French nuclear power stations shows that satisfactory solutions have been found in most cases. *

However, electricity-producing reactors which constitute centres in themselves have less complete processing installations than those forming part of a diversified complex possessing facilities designed to serve much wider needs than those of reactors alone.

At Marcoule, for example, the G-series reactors have the benefit of a set of almost complete facilities of the conventional type. Low- activity solid wastes are either incinerated or incorporated in concrete in 200-litre drums after compaction (if they are compressible). There is a pool for under-water cutting and an installation for cutting in air including, in particular, shears and blowpipes for reducing metallic pieces to suitable sizes. High-activity wastes are stored in trenches from which they are removed to join the normal processing cycle once their activity has decayed sufficiently. Liquid wastes are treated in a plant using either a simple filtration system, in the case of low-activity wastes (less than lO 3 Ci/m3 ), or chemical treatment, followed by incorporation of the sludges in bitumen, in the case of the most active wastes. Gaseous effluents are discharged into the atmosphere through suitable stacks; however, the carbon dioxide of the G2-G3 reactors is not discharged but liquefied, stored and re-used.

All these facilities are well suited for the purpose for which they were designed, i. e. to limit as much as possible any dispersal of radioactivity into the environment.

9 1


APPENDIX 3


APPENDIX 3

THE MANAGEMENT OF RADIOACTIVE WASTES

FROM COMMERCIAL NUCLEAR POWER STATIONS IN THE UNITED KINGDOM

I. DOUGALL Central Electricity Generating Board,

London,United Kingdom

INTRODUCTION

The first phase of the commercial nuclear power program in the United Kingdom comprises nine stations of the "magnox" type, with a total generating capacity of 5000 MW(e). In addition the United Kingdom Atomic Energy Authority prototype magnox stations at Calder Hall and Chapelcross together generate 376 MW(e).

In the second phase of the program a further 8000 MW(e) of nuclear power plant will be installed, of which the first three stations on order will have advanced gas-cooled reactors (AGRs). In these reactors the fuel is slightly enriched uranium dioxide canned in stainless steel, with graphite moderation and carbon dioxide coolant.

The reactors in both phases were constructed for the Central Electricity Generating Board (CEGB) and South of Scotland Electricity Board (SSEB) by industrial consortia with a background of design and operating experience from the United Kingdom Atomic Energy Authority's prototype magnox stations and WindscaleAGR (33 MW(e)).

Figure 1 shows the locations of these magnox and AGR stations, and Table I gives basic information for each station.

The present paper reviews the methods of waste management at the commercial magnox and AGR stations and includes data from operational experience at the magnox stations. The policies described and views expressed in the paper are in particular those of the CEGB, though in general the waste management policies of both the CEGB and SSEB are quite similar.

9 5 -


FIG. 1. Nuclear power station sites in the United Kingdom.

1. GENERAL DISCUSSION

1.1. Legislation

The present legislation regulating radioactive waste management at nuclear power reactors in the United Kingdom is as follows (Refs [1] and [2]:

9 6


TABLE I. BASIC INFORMATION ON COMMERCIAL UNITED KINGDOM NUCLEAR POWER STATIONS

StationElectrical

output(MW)

Type OperatedCommenced operation

by: Reactor 1 Reactor 2

Berkeley 276 Magnox steel pressure vessel

CEGB June 1962 Oct. 1962

Bradwell 300 Magnox steel pressure vessel

CEGB July 1962 Nov. 1962

Hunterston"A"

360 Magnox steel pressure vessel

SSEB Feb. 1964 July 1964

Hinkley Point "A"


CEGB March 1965 May 1965

Trawsfynydd 500 Magnox steel pressure vessel

CEGB March 1965 May 1965

Dungeness"A ”


CEGB Oct. 1965 Dec. 1965

Sizewell 580 Magnox steel pressure vessel

CEGB March 1966 Sept. 1966

Oldbury-on-Severn

600 Magnox concrete pressure vessel

CEGB Nov. 1967 Dec. 1967

Wylfa 1180 Magnox concrete pressure vessel

CEGB 1969 a

Dungeness"B"

1200 AGR CEGB 1970 a

Hinkley Point „ B„

1250 AGR CEGB 1972 a

Hunterston"B"

1250 AGR SSEB 1973 a

a Estimated

The Nuclear Installations Act 1965

This Act provides that nuclear power plants, other than those of the United Kingdom Atomic Energy Authority and government departments, shall only be operated under licence from the Ministries

97


of Power (or, in the case of Scotland, the Secretary of State). The Minister must attach to the nuclear site licence such conditions as he may think fit in the interests of safety; specifically these may include conditions relating to the handling, treatment and accumulation of radioactive waste.

The Radioactive Substances Act 1960

This Act regulates the keeping and use of radioactive material on premises other than those operated by the UKAEA or nuclear li censed sites. It also provides, however, that disposals of radioactive waste from nuclear power stations shall only be made following authorization by both the Minister of Housing and Local Government and the Minister of Agriculture, Fisheries and Food, or in the case of stations in Scotland, by the Secretary of State.

Applications of the above legislation

The practical application of the above legislation to nuclear- power-station waste management is summarized below. For the purposes of this document references to the "accumulation" of waste should be understood as out of process storage in a manner such that the waste is recoverable; "disposal" of waste may be by discharge from the licensed site or by burial on or off site, or in any case where there is no intent to recover the waste.

|a)_ Waste accumulation_

The submission to the Minister of Power (for his approval) of information in respect of each licensed site relating to the’ place and manner of accumulation of waste proposed at that site, and to the safety of the proposed accumulation.

^b)_ Wa_ste di£posal_

The submission, to the authorizing ministries of technical data on the nature of the wastes, their pre-treatment and the proposed methods of disposal.

Extensive consideration of safety aspects precedes the granting of all waste-disposal authorizations; this is illustrated by the work

9 8


of the Ministry of Agriculture, Fisheries and Food, on disposal of radioactive liquid wastes (see for example Ref. [3]).

Conditions generally attached to waste-disposal authorizations relate to:

(i) Sources of waste and quantities of radioactivity disposed(ii) Waste-sampling and monitoring methods(iii) Return of information on disposals made(iv) Environmental monitoring programs (off site, routine)

1.2. Basis of system design

The present methods of management of radioactive-waste arisings at nuclear power stations have developed from considerations of safety, siting and station layout, technical, legal and economic factors. These methods may be stated basically as:

(a) The long-term (i.e . possibly for a period beyond the station lifetime) accumulation of solid radioactive wastes (including solid residues from treatment of liquid and gaseous wastes).

(b) The disposal, after suitable treatment, of liquid and gaseous wastes, by discharge to the environment.

With these objectives, the general requirements of waste-system designs can be summarized as follows.

(a) Accumulation of waste:

(i) Location of facilities for safety, convenience and economy in waste handling and loading to stores.

(ii) Design of collection, treatment and handling equipment, and of accumulation facilities should be such that no significant radiological hazard can arise in operation.More specifically, account is taken of the need for; Shielding;Ventilation or provision of inert atmosphere;Drainage;Prevention of water ingress;Primary containment of wastes;Segregation of incompatible materials;Safety equipment, e .g . fire-detection instruments and

fire-fighting facilities;

9 9


Personnel access to stores in case of emergency; and Occasional remote viewing of store contents in normal

operation.(iii) Provision of adequate accumulation capacity. Specifically,

stores are designed to accommodate the estimated waste arisings during the operational life of the station or to be readily extendable. The estimated arisings are in general based on pessim istic assumptions as regards load factors, fuel-element irradiation, etc.

(iv) Stores are required to have a useful lifetime of at least 50 years, and to be so arranged that they can be sealed at the end of the station operating life, and thereafter will require a minimum of supervision.

(v) It must be possible to recover radioactive waste from accumulation facilities.

(b) Disposal of waste

(i) Reduction (by pre-treatment or otherwise) as far as practicable of the volume and radioactivity content of wastes discharged.

(ii) Generally as for (a) (ii) above.(iii) Arrangement of discharge-systems so as to achieve ade

quate dilution and dispersion of waste and thus to minimize the risk of radiological hazard to operating personnel and to members of the general public.

(iv) Provision of systems with adequate capacity to deal with maximum estimated arisings of waste.

(v) Provision for additional pre-treatment of wastes in the event of abnormal quantities of radioactivity being present in waste arisings.

Requirements (i), (ii) and (v) involve the application in design and operation, of local hydrological and meteorological data.

1.3. Capital and operational costs

Nuclear power stations are at present built for the CEGB and SSEB by industrial consortia, on a "turnkey" basis whereby the complete station is supplied under an overall contract; ancillary systems are not tendered for as separate entities and there is therefore no indication of the price of systems such as the waste-treatment and

1 0 0


accumulation facilities. It is not possible, therefore, to assess accurately the capital cost of providing facilities for waste systems. It is considered, however, that such costs would not exceed 1% of the total station capital cost.

In general, station costing procedures are such that waste - system costs are not readily separable from other operational and maintenance costs. Such information as is available, however, suggests that the annual cost of waste-system operation and maintenance is less than 1 % of the total station running and maintenance costs.

1.4. Operational experience

Quantities of wasteThe quantities of solid and liquid radioactive wastes arising from

the operation of the first six CEGB magnox stations and Hunterston "A" are summarized in Tables II, III and IV. Table V gives typical quantities of radioactivity discharged in gaseous wastes from three magnox stations.

Manpower and maintenanceIt has not been considered justified to maintain separate records

of manpower and maintenance requirements for radioactive-waste systems, and comprehensive discussion of this aspect is not given. However, the following observations may give some guidance.

Bradwell report that one plant attendant and two labourers are occupied full-tim e in the collection, accumulation or incineration of wastes from maintenance.

At Trawsfynydd, the operation of the pond water and active- effluent treatment plants occupies a chemist part-time, and two plant attendants full-tim e.

Hinkley Point mentions its manpower requirement of two man- hours per day for collection and assessment of gaseous-waste monitoring samples. This station comments that maintenance requirements for gaseous - waste monitoring systems are small and generally due to occasional failure of pumps and motors.

At one station the rotary magazine system for discharge of ir radiated fuel-element debris frequently locks, requiring maintenance, and modifications are in hand to eliminate the fault. Occasional maintenance is required on the active-effluent treatment plant at some stations because of deterioration of tank linings.

1 0 1


T A B L E I I . S O L I D - W A S T E A R I S I N G S A T M A G N O X S T A T I O N S

Station Berkeley15 Btadwell Hinkley Point"A” Trawsfynydd Sizewell Dungeness "A" Hunterston “A"

General category o f waste (1) (2) (3) (1) (2) (3) (1) (2). (3) (1) (2) (3) (1) (2) (3) (1) (2) (3) (1) (2) (3)

Fuel-element debrisc 15 000 60000 25 2400 8000 30 600 17 200 ~ 4 150 15 000 0 .8 N/A 6300 N/A N /A 26000 N /A 2000 (in 1967) 65000 5

Wet materials 360 ' 85 2800 3 NIL 18000 NIL -7 0 0 7630 9 NIL 2108 NIL N/A 1400 N/A N/A N/A N/A

Combustible wastes Negligible 60 000180 1200 15 NIL N/A NIL 30 10 750 Negligible. N /A Negligible

1-1920 3200 60 350 '

Non-combustibles^ 15 300 27 1312 4800 28 320 N/A N/A 6 10 540 480 N/A N/A 1.2000 •13000 ~12

Irradiated reactor components and instruments

~ 260 ~10 120 10 000 ~1 180 N/A ~1 10 N/A Negligible N/A N/A N/A•

a Excluding wastes in Table III. ( ! ) = Volume (ft*) o f waste produced to date,k Inventory also includes wastes from Berkeley Nuclear Laboratories. (2) = Total volume o f store (ft*).c Graphite sleeves, support members and zirconium bars. (3) = Percentage o f storage volume used to date.d Includes magnox from fuel elements.N/A = Information Not Available


TABLE III. ARISINGS OF LOW-ACTIVITY SOLID COMBUSTIBLE WASTE AT MAGNOX NUCLEAR POWER STATIONS

Station Bradwell Berkeley Hinkley Point "A" Trawsfynydd Sizewell Dungeness "A" Oldbury a Hunterston"A'

Categoryb "A" lb/yr 3000 . 12 000 32 000 6400 4200 3300 12 000 12 000

Category'3 ”B" lb/yr 39 000 24 000 38 000 109 000 42 000 10 000 36 000 N/A

Category'3 "C" lb/yr 36 000 36000 87 000 64 000 210 000 - N/A

Assumption: Average density of paper waste as drummed = 3.5 lb/ft.3

a Pre-operational estimate.

k Category "A " = Wastes from contamination zones.

Category "B" = Wastes from within Reactor Area but not from contamination zones.

Category ”C” = Wastes from outside Reactor Area.


TABLE IV. VOLUMES OF ACTIVE LIQUID-WASTE ARISINGS AT MAGNOX STATIONS AND QUANTITIES OF RADIOACTIVITY DISCHARGED

Typical volume

arisings (gal/yr)

Year and activity (Ci) discharged

Station 1962 1963 1964 1965 1966 1967

(1) (2) (1) (2) (1) (2) (1) (2) (1) (2) (1) (2)

Berkeley (a) 2.4 x 10s(b) 10

0.003 65 1062 4.7 306 23 225 70 311

Bradwell (a) l.O x 10s(b) 290 0.006 NIL 0.2 233 4.1 37 19 471 29 583 100 499

Hunterston "A" (a) 1.6 x 107 (1) (2) (1) (2) (1) (2) (1) (2)(b) 320 <0.02 21.2 1.6 '477 18.8 477 19.9 126

Hinkley Point "A" (a) 3.6 X 106(b) 15

0.001 0.006 NIL 2 174 18 32 7.9 NIL

Trawsfynydd (a) 3.0 X 10s(b) 176

0.053 2.1 2.2 221 7.1 110

Dungeness "A" (a) 1.9 x 10s(b) 770 0.01 6.0 2.3 569 40 528

Sizewell (a) 2 .4 x 107(b) 31 1.0 29 2.5 116

Oldbury (a) -(b) -

0.016 3.7

N.B. Where only one figure is shown more than 99% of activity is due to tritium;otherwise (1) = Activity other than tritium. (a) = Arisings from active effluent treatment plant.

(2) = Tritium. (b) = Arisings from humidriers.


TABLE V. MEAN ANNUAL DISCHARGES OF PARTICULATE RADIOACTIVITY IN GASEOUS EFFLUENTS FROM THREE MAGNOX STATIONS

Source Shield-cooling aira (mCi/reactor per yr)

Reactor COz coolant (mCi/reactor per yr)Station

Bradwell 2.2 0.07

Hinkley Point "A" 16.3 1.04

Sizewell 8.7 0.72

a The emission rate of 4JA activity in shield-cooling air at CEGB Stations is from10-15 Ci/h per reactor.

Unusual incidents

Operation and maintenance work at nuclear power stations, including waste handling, is carried out in accordance with procedures specified in station operating instructions and manuals. Basic safety standards are specified in the Safety Rules (Radiological) and in the nuclear site licence conditions (see Section 1. 1); the maximum permissible levels of radiation and contamination are in conformity with, and in some cases lower than,the ICRP standards currently re commended.

Whilst there have been no serious incidents to date involving radioactive-waste management, a number of minor mishaps have occurred. Two such incidents are briefly described:

(1) An accidental release of airborne contamination occurred at Bradwell Power Station in 1965, during the emptying of dust from the bypass filter pots of the No. 1 reactor. (A proportion of the re actor gas is continuously fed through these filters to control dust levels in the circuit.)

Standard procedure for this operation requires the establishment of zero pressure by valve adjustment in the filter, the filter-pot securing bolts then being loosened and the filter pot removed. P rotective clothing is worn for this operation. On the occasion in ques

1 0 5


tion, a residual pressure existed in the filter pot which caused ejection of dust when the pot was loosened. Some dust escaped from the room, resulting in slight contamination of personnel in adjacent areas.

The operation was complicated by the location in separate areas of filter pressure gauges and valve controls, respectively. These have since been grouped together outside the filter room and, in addition, inter-communicationfacilities provided between the pressure- control operator outside the room and the operator inside it.

(2) At Trawsfynydd Power Station various irradiated metal components from the reactors, including flux-flattening bars, are a ccumulated in an underground active-waste vault within the building.

The flux-flattening bars are normally discharged from the re actor charge face (using the charge machine), and fall into the vault via a connecting tube, which itself passes through a room (normally locked) allowing access to a boiler gas duct. During May 1967, on an occasion when entry to the room was required for duct inspection, a high radiation field was detected. The source of radiation proved to be a blockage of flux-flattening bars and trailing lead thermocouple wire (also discharged from charge face) within the disposal tube, the radiation dose-rate on the-outside of which was 110 R/h.

The blockage was cleared by a special operation carried out from the boiler-duct access room. A section of the disposal tube was cut away, the blockage cleared using tongs, temporary shielding, e tc ., and the tube rewelded. The highest personnel radiation exposure was 0. 52 rem (whole-body dose).

During the period of maintenance on the disposal tube flux- flattening bar disposals to the vault were made via an alternative access route normally used for other types of waste.

Changes in reactor operation

The volumes and activity levels of some waste arisings are dependent, sometimes to a considerable extent, upon reactor operating conditions. For example:

(i) The achievement of fuel-element irradiation times in excess of design targets reduces element throughput rates and consequently the volumes of fuel-element debris for accumulation. Levels of radioactivity of gaseous andli-

1 0 6


quid wastes may, however, increase in some circum stances, owing to the higher probability of fuel-element magnox-can penetration while in the reactor and subsequently in the fuel-element cooling pond. (See Section3.)

(ii) Operation of reactors with relatively high moisture levels in the pressure circuit, due for example to boiler leakage, results in increased arisings of liquors from the gas- drying plant. Present evidence indicates that the specific activity of tritium in these liquors is relatively independent of the volumes arising; thus, the quantities of tritium for disposal are closely related to the gas-circuit moisture levels.

(iii) The levels of induced gaseous and particulate activity in shield cooling air discharges at steel pressure-vessel stations (see Section 4) are directly dependent on reactor power levels, among other factors.

(iv) At some stations, problems of oil ingress from the shaft oil seals on the gas circulators have been encountered. A system of controlled purging of coolant gas through the seals has been adopted by some stations to minimize this ingress of oil; the purge gas constitutes an additional source of radioactive gaseous waste.

(v) Variations in methods of reactor control may remove the requirement for components of the original control system and thus create wastes. Thus, at one station the proposed use of control rods, instead of flux-flattening elements, for flux "trimming" is likely to necessitate the disposal of a large number of these elements.

(vi) Miscellaneous combustible and non-combustible wastes of very low radioactivity content are produced in relatively large quantities during reactor maintenance periods.

Recommended changes in original designs

(1) Several stations commented, at an early stage of operation, on the lack or inadequacy of space for:

(a) Temporary accumulation of combustible solid wastes pending transport off site.

(b) Buffer storage of miscellaneous active waste for sorting prior to accumulation on site.

1 0 7


108 T A B L E V I. PROVISIONS FOR ACC U M U LA TIO N O F SO L ID -W A ST E ARISINGS

(i) B ra d w ell

S a fe ty features

Store D escrip t ion L oca tion M a teria ls storedV o lu m e

a l lo c a te d( f t 3)

E xp ectedarisings

( f t 3> V e n t ila t io n D ra in a geM eth od o f loa d in g

w astes to store

F ire -d e te c t io n and f ir e - f igh tin g

prov ision s

1 . A c t iv e w aste U nderground c o n c r e t e A d ja c e n t to F u e l-e le m e n t 8000 Not Y es Y es S h ie ld ed flask Yes

dum p va u lt w ith r o o f at ground l e v e l . 12 c e l ls ,

fu e l-e le m e n c c o o l in g pond

debris ( 4 c e l l s ) a v a ila b le (co n tin u o u s ) r o o f a c c e ss

to ta l storage v o lu m e 19, W et m a ter ia ls 27 90 N ot N o Y es Pum ped as slurry N o

190 f t 1 ( 3 ce lls ) a v a ila b le s u p em a ten t liq u or d e ca n te d a fte r s e tt le m e n t

S h ie ld in g : N o n -co m b u s t ib le ( 4 ce lls )

6 3 00 N o Y es R o o f a c c e s s using o v e rh e a d gantry

N o

R o o f: 2 ft 6 in . c ra n e (5 t)c o n c r e t e

O u ter w a lls : 1 ft 6 in . C om b u stib le s 2 1 0 0 N o Y es R oo f a c c e s s using Y esc o n c r e t e (1 c e l l ) ov erh ea d gantry

c ra n e (-5 t)

2 . P i le -c a p S te e l tubes suspended R eactor Long rea c to r 2 4 tubes N ot N o N o From p ile ca p N oa c c u m u la t io n from ch a rg e f a c e , b e b u ild in g c om p on en ts , e . g . per rea c to r a v a ila b le using ch a rg e

tubes tw een re a c to r b i o l o g ic a l sh ie ld in g . One s e t 'o f 24 per r e a c to r .

c o n tro l rods m a ch in e

3 . M a in ten a n ce V o id b en ea th low er R eactor M isce lla n eou s 60 00 N ot N o N o From m a in te n a n c e No

c e l l v o id s r em ote m a in ten a n ce c e l l . O n e v o id per r e a c to r . V o lu m e ~ 6 0 0 0 f t 3

bu ild in g rea c to r c o m ponents d is ca rd ed during m a in ten a n ce

(e a c h ) a v a ila b le c e l l


109

TABLE VI (cont.)

(ii) Wylfa

Store D escr ip t ion L oca tion M a ter ia ls stored

E xp ected arisings

in 30 yr ( f t 3 )

S a fe ty features

V e n t ila t io n D ra in a geM eth od o f

lo a d in g w astes to store

F ire -d e te c t io n and f ire f ig h tin g

prov isions

1 . M ain debris d isp osa l v o id (n orth )

S h ie ld e d v o id v o lu m e 3 7 4 0 0 f t 3

C en tra l b lo ck sh ie ld ed fa c i l i t ie s

S m a ll rea c to r c o m p o n e n ts ( f lu x - fla tten in g b a n )

A c t iv e debris from fu e l route

2 1 0 0

200

No

(A ir " c u r ta in " is m a in ta in ed a cross a cce ss o p en in g s to v o id )

Y es D ir e ct d isch arge from ch a rg e fa c e by ch a rg e or s e r v ice m a ch in es

N o

2 . M ain debris v o id (sou th )

S h ie ld e d v o id v o lu m e 3 4 0 0 0 f t 3

C en tra l b lo c k sh ie ld ed fa c i l i t ie s

As a b o v e plus C 0 2 c ir c u it filters and C O jd rie r dess ica n t a

900

1 0 00 aAs a b o v e Y es A s a b o v e o r from

ch a rg e f a c e by hand

N o

3 . R e m o te -h a n d lin g fa c i l i t i e s v o id

S h ie ld e d v o id b e lo w r e m o te -h a n d lin g f a c i l i t i e s . V o lu m e 5 3 1 0 0 f t 3

C en tra l b lo ck sh ie ld ed fa c i l i t ie s

R ea cto r co m p o n e n ts co n t ro l rods

Others

96 rods

39 70

Y es Y es Jettison ed from RHF

No-

4 . C o m b u stib le w aste store

V o lu m e 2 1 0 0 f t 3 C en tra l b lo ck sh ie ld ed fa c i l i t ie s

C om b u stib le wastes - Y es Y e s - v ia a d ja c e n t area

M an u al loa d in g Y es

5 . S lud ge storage tanks (3 )

V o lu m e 20 55 0 f t 3 C en tra l b lo ck sh ie ld ed fa c i l i t ie s

W et m a te r ia ls C 0 2 p re ss u re -c ir cu it dusts

14 700

300

V en ts to r o o f le v e i

Y e s (p lu sle a k -d e te c t i o rfa c i l i t ie s )

Slurried o r f lu id iz e d and pu m ped to tanks. Supernatent liq uors d eca n ted a fte r se tt le m e n t

No

a Figures v ery a p p rox im a te - for illustra tive purposes o n ly


110

TABLE VI. (cont.)

(iii) Dungeness "B "

Store D escrip tion L oca tion M a ter ia ls stored

E xpected arisings in 30 yr

( f t 3 )

S a fe ty features

V e n t ila t io n D rainageM eth od o f

lo a d in g wastes to store

F ir e -d e te c t io n and f ir e - f igh tin g

p rov ision s

1 . F u e l-e le m e n r debris v o id

V o lu m e = 70 0 0 0 f t3 S tee l l in es , C 0 2 f i l le d sh ie ld ed v o id

B elow fu e l- h a n d lin g u n it, c e n t ra l b lo c k

F uel stringer co m p o n e n ts

2 4 000 N o N o , but p r o - p rov is ion for d ra in a ge in s p e c ia lc ir c u m sta n c e s

D rop p in g from fu e l h a n d lin g w orkshop

Y es

2 . S e r v ic e ba y v a u lt

V o lu m e = 7-500 ft3 sh ie ld ed va u lt

B elow c o n t r o l-r o d se r v ice bay, ce n t ra l b lo ck

R ea ctor c o m p o n e n ts , c o n t ro l rods

Others

69 rods

515

Y es Y es D rop p in g from c o n t r o l - r o d s e r v ice bay

N o

3 . M a in w aste v a u lt

V o lu m e = 1 0 000 f t3 sh ie ld ed va u lt

C o m m o n serv ices b u ild in g near a c t iv e w orkshops and d e co n ta m in a tion c en tre

G a s -c ir c u it f ilters and o th e r n on co m b u stib le s

540 Y e s Y es T o w e re d by c ra n e

N o

4 . F i lte r -h o ld in g v a u lt

V o lu m e = 30 000 f t 3 V a u lt s h ie ld e d , but for structural reasons o n ly

C o m m o n serv ices b u ild in g

H ea tin g and v e n t ila t io n filters

11 000 Y es Y es By hand Y es

5 . D ess ica n t vau lt V o lu m e = 1 2 00 0 f t 3 V a u lt s h ie ld e d , but for structural reasons o n ly


COz d r ie r d ess ica n t 10 000 Y es Y es By hand N o

6 . S lud ge storage tanks (2 )

M ild s tee l c y l in d r ic a l tanks. T h ick n e ss : B ottom 3 / 4 in .S ides 5 /8 in .T o p 5 /1 6 in . S h ie ld ed by 2 - 3 - f t

co n c re te


W et m a ter ia ls N /A T anks v en ted and area v e n t ila te d

Y es W astes s lurried and p u m p ed to tanks su p em atent liq u ors d e ca n te d

N o


(2) At nearly all stations, a high proportion (90 - 95%) of miscellaneous waste arisings (i.e . excluding fuel-element debris, wet materials and irradiated reactor components) have surface radiation dose-rates of less than 10 mR/h. The accumulation of such wastes in highly shielded facilities (see Table VI) is expensive, and in some cases the volumes arising are such that these facilities may be filled prematurely; the provision of more spacious and lightly shielded stores for these wastes is therefore a foreseeable requirement at some stations.

(3) The quantities of miscellaneous combustible waste arising per year at magnox stations are summarized in Table III. A large proportion (categories B and C) of these wastes are inactive or only nominally radioactive, and a scheme whereby each nuclear station would have a small incinerator for disposal of these wastes is under consideration.

(4) At Bradwell Power Station, the operation of the active-waste incinerator unit (see Solid Wastes) has been considerably hampered by the need for frequent changing of the "absolute" filters, which quickly become clogged in operation. In view of the very low activity content of the-wastes burned in the incinerator, permission has been given for a trial period of operation of this unit without absolute filters.

(5) Other comments on limitations of design, applicable to only one or two stations and in no case causing any serious difficulty or hazard to personnel, include:

(i) Difficulties with plant for the treatment (de-watering) of filter sludges prior to accumulation.

(ii) Failure of tank linings in plant for active liquid-waste treatment.

(iii) Inadequate capacity, and overshielding of fuel-element debris transfer flasks (see Solid Wastes).

(iv) At one station, the lack of adequate drainage from the concrete covers of the active-waste vaults, together with difficulty in maintaining good seals on the covers, has re sulted in ingress of water to the vault.

I l l


2. SOLID WASTES

The solid wastes arising from magnox and AGR operation are divided into five categories for description of methods of collection, treatment and accumulation. The categories are:

(i) Fuel-element debris(ii) Wet materials(iii) Combustible wastes(iv) Non-combustible wastes(v) Irradiated metallic reactor componentsTable VI summarizes the provisions for accumulation of solid-

waste arisings at Bradwell, Wylfa and Dungeness "D" andalsoquotes the estimated arisings for the last two stations. Table VII indicates the nature and the order of activity content of some magnox station wastes in each category.

Owing to the varied nature and activity levels of the wastes and to the various means adopted at different stations for their handling and accumulation, the above division is rather arbitrary. The categories represent the general segregation system at most magnox stations but are not entirely applicable at Sizewell and Wylfa or for the present AGR stations.

A notable trend in waste management is towards the reduction or elimination of manual handling of potentially hazardous wastes rby arranging for their direct discharge from source to store.

2.1. Sources and types

2. 1. 1. Fuel-element debris

.LaL Magnox_ s_tations_

After removal from the reactors, irradiated magnox fuel e lements are discharged, at stations other than Wylfa, to water-filled ponds for storage to allow short-lived fission products to decay, before transporting the elements to Windscale for chemical processing. Prior to element transport off site, certain external components (mainly splitter vanes) are removed by special machinery. Figure 2 shows a typical polyzonal magnox fuel element and indicates the parts removed. In general the debris produced by "desplittering" or "delugging" consists mainly of magnox.

1 1 2


FIG. 2. Magnox polyzonal fuel element.

It may be noted that no waste of this type will be accumulated at Wylfabecause of the special arrangements (including dry fuel storage) at this station.

£b)_ AGR_stations _

The AGR fuel elements will consist of fuel pans (uranium oxide canned in stainless steel) in a steel lattice sleeved in graphite. The

1 1 3


114 TABLE VII. SOME TYPICAL ACTIVITY LEVELS OF SOLID WASTES IN MAGNOX STATIONS AND NATURE OF THE ACTIVITY a

Category of waste Item Typical initial activity Nature of activity

Fuel-elementdebris

Magnox splitters (magnox stations)

110 /iCi/g (8 Ci/ft3)

plus 1 mCi/g

Mainly 51Cr, 55/ 59Fe 65Zn induced by irradiation of magnox24Na from Mg irradiation

Wet materials

Cooling-pond sludge

Spent sands and resins Pressure-vessel cooling Water-treatment sludge

250 mCi/ftS

Very low 0.2 Ci (60Co)/ft3

Insoluble magnox corrosion products including 51Cr, 55/ 59Fe, 6SZ.n and “ Co.

60Co,55Fe and56Mn from steel pipes

Non-combustibles

CO 2 drier alumina COj circuit filter dust

Low High - say 100 mCi/g

Compounds containing tritium Activation products of structural materials, e.g. s^e,'s5Fe, 54 Mn and60Co in oxides

Combustibles

Waste paper, clothing, etc.

Shield-cooling-air filters

Low - a few millicuries per drum

Low - a few millicuries per filter

Depends on source of waste

Activation products of concrete and steel, e.g. “ Fe, 55Fe, MMn and iMSb


TA B LE VII (cont.)

Category of waste Item Typical initial activity Nature of activity

Irradiated metal components from reactors

Flux-flattening bars Control rods Charge chutes Charge-machine grabheads

200 Ci/unit 2000 Ci/unit 300 Ci/unit 200 Ci/unit -

Activation products of steel and impurities 54Mn, s6Mn, 60Co, 55Fe, 59Fe

a Figures are very approximate and for illustrative purposes only


elements will be assembled into stringers with a connecting metal tie rodandtop- and bottom-end assemblies. Irradiated fuel stringers will be dismantled (prior to element decay storage) in dry shielded cells and the debris so produced will consist of graphite and metal from stringer components other than the fuel elements.

2 .1 .2 . Wet materials

The wastes in this category com prise mainly pond and filter sludges, spent sands and ion-exchange materials. These arise from the treatment systems (see Section 3) for fuel-element cooling pond, active-liquid effluents and concrete pressure-vessel cooling water, respectively. Sludges from decontamination processes are also included.

2. 1.3. Combustible wastes

The wastes in this category include:

Ventilation plant filters Protective clothingTemporary protective floor coverings SwabsShield cooling air filters (some steel pressure-vessel stations) Paper handkerchiefs and towels.

The rate of arising of these wastes depends on factors such as maintenance requirements and spillages and is therefore likely to be very variable.

2 .1 .4 . Non-combustible waste

Wastes in this category include the following:

Filter elements from COz circuit blowdowns, bypass and relief valve systems

Dusts from C 0 2 circuit filtersDebris from COz ducts and reactor pressure vessel Dessicant from coolant gas drying plant Ventilation filtersShield cooling air filters (some steel pressure-vessel stations) Ash from the incineration of low-activity combustible wastes.

1 1 6


As in the case of combustible wastes, accurate estimates of certain arisings, e .g . dessicant, are very difficult to make, especially in the case of the AGRs where there is insufficient operating experience from which to make extrapolation.

2. 1.5. Irradiated metallic reactor components

The wastes in this category are components of reactor fuelling and control systems which have become expended or damaged. These wastes arise either in the reactor or in the remote-handling facilities where maintenance work is done on highly active items. They include:

Control rods Flux-flattening bars Charge chutes Shield plugs GrabsTie rods (AGRs only)

2.2. Collection and treatment

2 .2 .1 . Fuel-element debris

Magnox_ s_tations_

The longer pieces of debris produced by desplittering are sometimes chopped into lengths of one or two inches and together with other small pieces of waste are dropped either directly into a shielded transport flask or into baskets or skips for subsequent loading into a transit flask. Once this has been done the flask is handled to the debris store and the waste discharged into the store for longterm accumulation. In the case of Trawsfynydd, however, the transfer of waste to the debris store is accomplished by a system of tubular vibrating conveyors. These move in shielded conveyor chambers over the store, which provide protection to personnel during the transfer operation.

At Hunterston "A" a large proportion of the solid-waste arisings consistsofgraphitesleev.es, support members and zirconium "D” bars from fuel elements. These components are separated from the fuel elements in dry shielded cells and passed to a "cracker"

1 1 7


unit for sleeve crushing. They are then transferred to accumulation bunkers in a building adjacent to the reactors by a remotely controlled system, which includes a shielded railway, skip hoist and dumping room. Magnox components are removed in the cooling pond and are baled prior to storage.

All plant designs include interlocks to provide protection against:

(i) Inclusion of irradiated fuel in debris introduced to store(ii) Discharge of debris while en route to store(iii) Exposure of personnel to radiation or contamination during

transfer and discharge of debris(iv) Discharge of fuel element debris to a store other than that

designated for debris accumulation.

£b)_ AGR_stations_

The debris from the dismantling of the fuel stringers in the ir radiated fuel dismantling cells will be discharged directly to a shielded void below the cell; this removes the need for provision of interlocks such as (a) (ii) - (iv) above. Interlocks to prevent the inadvertent discharge of a fuel element to the debris void are to be provided.

A comparison of systems (a) and (b) demonstrates the trend in the latest stations towards direct discharge of waste from the place of arising to the store.


Filter sludges are backwashed to settling tanks using installed pumps and pipework. From these tanks the sludges pass either directly or via the de-watering plant (drum drier or pressure filter) to shielded storage tanks. Where there is no pre-storage treatment, the sludge is de-watered after settlement in the tank by pumping off the supernatent liquors.

Spent filter and ion-exchange materials are stored with filter sludges or in an adjacent tank. At most stations permanent pumps or ejectors, and pipework are installed for sand and resin transfer operations.

1 1 8


2 .2 .3 . Combustible wastes

Most of these wastes have a very low radioactivity content. They are manually collected at source, wrapped and then metal-contained for accumulation.

2 .2 .4 . Non-combustible wastes

Varipus procedures are adopted at present stations for collection and treatment, depending on the specific nature of the wastes and how they will be accumulated.

Wastes such as pressure-circuit-filter dusts may have associated high-activity levels, and careful radiological control is required in collection operations. Filter-dust-collection pots are usually integral with the filter units; the pots are detached for transfer and permanent accumulation, or for emptying by fluidizing the dust and discharging it to wet stores (e.g. sludge tanks), and re-use. At one new station, however, the filter dust catch pots will themselves constitute the accumulation facilities. They have capacity for the estimated dust arisings during station lifetime and will remain in situ beneath the circuit filters.

The trend in handling of spent reactor gas-drier dessicant is also away from manual or vacuum collection in small transfer/accumulation drums, towards direct discharge from the drier beds to store.

2 .2 .5 . Irradiated reactor components

Wastes arising in the reactor are handled by charge-face machinery, including the fuelling machines, and discharged to the accumulation facility via access points on the charge face. Where the wastes arise in the remote maintenance cells they are discharged direct to store from the cells.

Apart from drum winding of wires, or chopping of hoses into shorter lengths to facilitate discharge, pre-storage treatment of these wastes is minimal.

Various design and operational controls are adopted at the magnox stations to ensure that no incompatible materials, in particular irradiated fuel elements, are discharged to the facilities. Design interlocks are variously of a mechanical, (e .g . constriction in the access tube to the void to prevent passage of an element) or e lec

119


trical nature (e .g . location of electrical supply point such that the charge-machine fuel-element hoist is inoperable over the discharge hole).

In the case of discharges to stores via remote handling facilities, the direct discharge of elements from the fuelling machine is usually prevented by off-setting the access holes and by grab "depth" interlocks (such that the fuelling-machine grab is inoperative in the facility at the operator working level).

In addition to any design provisions made in this respect, strict administrative controls apply to all operations involving charge-face machinery. The risk of irradiated fuel discharge to any waste- accumulation facility in the reactor is therefore small.

2 .3. Accumulation of solid wastes

2 .3 .1 . Fuel-element debris

(el)_ Magnox s_tatiqns_

Because of the pyrophoric properties of finely divided magnox and the evolution of hydrogen from the magnesium/water reaction, magnox is segregated from other wastes and special safety precautions are incorporated in the design of all magnox debris stores. At two stations, underwater storage was adopted in order to minimize the risk of fire. Present experience suggests, however, that accumulation under dry conditions is entirely safe. The design of desplittering apparatus prevents the inclusion in the debris of finely divided material and adequate ventilation ensures that hydrogen concentrations in the store are maintained at safe levels. Dry debris stores, in addition to having adequate capacity and shielding, incorporate provisions to ensure that:

(i) Ingress of ground or rain-water is small and contact of such water with the debris is avoided (e .g . by provision of a layer of gravel on the cell floor).

(ii) Drainage of the facilities is adequate.(iii) Ventilation is adequate. Ideally the air should be intro

duced across the base of the cell so that it sweeps upwards through the debris mass.

(iv) There is temperature monitoring in the debris mass together with monitoring of the air temperature in the store (with provision for alarm on the basis of high temperature

1 2 0


and high rate of temperature r ise ). These instruments are arranged to alarm to continuously manned areas.

(v) The cell atmosphere can be sampled for hydrogen content measurement.

(vi) Arrangements exist for the introduction of water for debris cooling.

Figure 3 includes a typical dry store for magnox debris.

£b)_ AGR_statiLons _

AGR fuel stringer debris, consisting mainly of graphite with smaller quantities of steel, will be accumulateid in dry shielded vaults in the central services block between the reactors. A typical arrangement is shown in F ig .4.

The detailed safety requirements of debris vault design are at present under consideration. At one station provision has been made for debris storage in a CO2 atmosphere to cater for possible fire risk associated with the presence of graphite dust deposits on the debris.


Sludges and spent sands and resins are accumulated on site in shielded stores close to or integral with the pond water and active - effluent treatment plants. At three early magnox stations, these wastes are stored in drums. However, the preferred method of accumulation is either in large mild steel tanks contained in concrete vaults or cells (see Fig. 3 ) or, if storage above ground is adopted, in shielded concrete tanks erected on an impervious tray.

Design requirements for such facilities include the following:(i) Mild steel tanks should be lined internally, e .g . with a

multi-coat neoprene lining not less than 0.050 in. thick, and painted externally, or alternatively the thickness may include a corrosion allowance. Concrete tanks may have an internal lining of asphalt or equivalent material.

(ii) Leakage detection arrangements must be provided.(iii) Where mild steel tanks are used, some provision should

be made for chemical dosing of the slurry liquors to an alkaline condition, in order to minimize tank corrosion.

1 2 1


122

FIG. 3. Typical arrangement of a general active waste dump.


NEW & IRRADIATED FUEL H A N D LIN G ROOM

FUEL BUFFER STORAGE

FIG.4. Longitudinal section of pre- and post-irradiated fuel handling unit.


(iv) At least two tanks should be provided for the estimated station lifetime arisings. In the event that repair work was necessary on one tank it should be possible to pump the contents into the other tank or into shielded drums.

(v) Tank vents to atmosphere should incorporate filters to prevent the release of particulate activity, or alternatively should discharge at high level.

2 .3 .3 . Combustible wastes

After collection and containment in polythene (or PVC wrapping inside light-gauge metal containers) these wastes are segregated at source or in a sorting facility into two sub-categories, viz.:

(i) Wastes which, by reason of high activity content, associated non-combustible components or otherwise, cannot be incinerated.

(ii) Wastes for incineration.The wastes in (i) are accumulated on site in shielded facilities,

common features of which are:(a) Fire-detection instruments and alarms (to a continuously

manned area) and fire-fighting facilities.(b) Means of preventing rain-water from entering store.(c) Means (e.g . collection sump) for detecting ingress of

water and for discharging it to the active drainage system (see Liquid Wastes).

(d) An extract ventilation system with filtered discharge and means of sampling the discharge.

(e) Adequate loading arrangements to ensure that waste containers will not be damaged in loading to stores and that reasonable packing factors can be achieved.

Wastes in (ii) are accumulated in transit stores pending despatch (by road) from the site for incineration (see Section 2. 3.4 below). The containers used comply with requirements of IAEA Transport Regulations for Type "A" containers and their activity content (generally estimated from radiation levels at the container surface) is within the appropriate limits. The transit stores are situated within the reactor areas at places convenient for loading to transport vehicles. The stores are well ventilated, preferably weatherproofed, and provided with fire-detection and fire-fighting facilities.

124


2.3. 4. Disposal by incineration

The CEGB at present has two active-waste incinerators which are located at Bradwell Power Station and Berkeley Nuclear Laboratories (BNL is on the same site as Berkeley Power Station), respectively. These units were installed to cope with arisings of low- activity combustible wastes. Their location was arranged so that Bradwell normally burns its own wastes together with those from Dungeness "A", Sizewell and Trawsfynydd, whereas wastes from Berkeley, Hinkley Point "A" and Oldbury are burned at BNL. The installation of a third active incinerator unit, at Trawsfynydd, is being considered.

The ash from incineration is accumulated by the incineratingsite.

At Bradwell the incinerator off-gases are cleaned by cyclone and absolute filters {but see Operating Experience). Treatment in the BNL unit consists of water scrubbing, and filtration by absolute filters. A schematic of this system is given in Fig. 5. Both units employ after-burners.

A S H C A N W A T E R T O T R E A T M E N TP L A N T

F I G . 5 . S c b e m a t i c a r r a n g e m e n t o f a c t i v e w a s t e i n c i n e r a t o r .

1 2 5


TA B LE VIII. SOME DATA FROM OPERATION OF ACTIVE-W ASTE INCINERATORS

Incinerator Bradwell BNL Hunterston "A"

Period to which data relates

1967 Nov. 1965 - Dec. 1966 1967

Weight of waste burned 40 000 lb 85 120 lb 2600 ft3(volume burned)

Weight of ash produced 1000 lb 4816 lb -

Volume (approx.) of ash a produced

72 ft3 344 ft 3 120 ft 3

Volume of spent filters produced

360 ft3 135 ft 3 NIL

Controls applied to ♦ 7.5 mR/h at surface of £ 15 mR/h at surface of any < 3mCi of 35S + < lmCi of other nuclideswaste burned any waste container waste container in waste consigned per month

a Ash density - 14 lb/fts.


"Hunterston report satisfactory experience with their small incinerator for the disposal of very-low-activity wastes and results with this unit are included in Table VIII. The provision of a larger incinerator for dealing with more active wastes at this station is being considered."

Table VIII gives some results from the operation of these units.

2 .3.5. Non-combustible wastesThese wastes are stored in shielded cells or voids the design

requirements for which are the same as those for combustible wastes except that fire-detection and fire-fighting provisions are not required.

Container requirements for accumulation in stores depend largely on the type of store and the degree of segregation of incompatible materials. Where these materials are stored collectively in dry stores, typical container standards are:

(i) Dust-loaded wastes (other than (ii)) with surface-radiation levels of 7.5 mR/h or more, are polythene wrapped and placed in light-gauge metal drums or boxes.

(ii) Humidrier dessicant - as for (i).(iii) Pressure-circuit filters and dusts are polythene wrapped

and placed in air-tight steel containers of thickness not less than 1/10 in. and 1/ 5 in ., respectively.

(iv) Miscellaneous low-active materials are bagged in polythene where practicable.

2 .3.6. Irradiated metal components

Because of the high radiation levels from these wastes and of handling considerations they are accumulated at all stations, except Berkeley, in shielded tubes, vaults or voids within the reactor buildings.

At the earlier magnox stations (before Sizewell), where the re actors are housed in separate buildings, the storage tubes and voids are built into the reactor biological shields (primary and secondary). At later stations, with the integration of the reactor buildings, corresponding facilities are located in the common central services block beneath the shielded remote maintenance facilities and (in the AGRs) fuel-dismantling cells.

Figure 4 illustrates the arrangement of typical storage facilities for these wastes at an AGR station.

127


3.1. Sources, types and quantities

3. LIQUID WASTES

The main sources of radioactive liquid wastes at magnox and AGR stations are:

(1) Fuel-element cooling ponds.(2) Reactor pressure-vessel coolirig systems (concrete

pressure-vessel stations only).(3) Reactor-coolant drying plant.(4) Other sources of relatively low-activity liquors. Typical volumes arising from these sources are given in

Table IX.

3.1.1. Fuel-element cooling ponds

|a)_ Magnox_ s_tations_

With the exception of Wylfa, all the magnox stations store the irradiated fuel elements in cooling ponds prior to transport off-site for chemical processing. Fuel-element dwell-times in the ponds depend on irradiation history and on other factors such as transport conditions, but are typically 70 to 130 d.

To minimize corrosion of the element (magnox) cans and to control the levels of soluble and particulate radioactivity in the ponds of CEGB stations, treatment plants are, provided with incorporating coolers, filters and ion-exchange materials together with degassing and chemical dosing plant. The liquid wastes from these plants consist of filter backwash liquors (aqueous) and ion- exchange regenerant liquors (acid and alkaline) together with large volumes from the occasional emptying of pond sections for maintenance.

Figure 6 shows a typical pond-water treatment scheme and Fig.7 shows an additional plant, which has been installed at stations where activity discharge limits are especially low.

"At the SSEB's Hunterston "A" station, pond conditions are controlled by chemical treatment and continuous purging of pond-water via a precipitator unit and sand pressure filters ."

128


TABLE IX. SOURCES AND TYPICAL ESTIMATED MAXIMUM VOLUMES OF LIQUID WASTE FROM A MAGNOX AND AGR STATION

Source

Volumes arising (gal/d)

Magnox station (500 MW(e))

AGR station (1200 MW(e))

Changing rooms and laundry 3000

1 3000

(no laundry at "B" station)

Charge-machine washdown Decontamination centre Transport-flask washdown

j- 1500

Ion-exchange-plantregenerants 3000 NIL

Filter backwash 500 500

Pond purge NIL 1400 - 2800

Humidrier liquors

Other sources. small

60

small

£b)_ AG_R_statiLons

The Dungeness "B" and Hinkley Point "B" stations will also have cooling ponds for element decay storage. Because of the different properties of the element can material, however, no re quirement for a chemical treatment or ion-exchange plant is at pre-

129


P O N D PUMPS COOLERS PUMPS FILTR ATIO NSECTIONS

C A T IO N UNITSPLANT

SPACE FORC O 2 IN JE C T IO N TOWER ( %£ f c A No CO LUM N &SEQUESTERING UNIT

r O i

D E G A S I F I E R P U M P S M I X E D BED / n UNIT

h Uf m

L @ J © 0 K -CAUSTIC IN JE C T IO N

EMERGENCY C O N N E C T IO N FROM ACTIVE EFFLUENT TREATMENT PLANT 0 —

NOTE:FILTER SLUDGE & IO N EX CH A N GE CO LUM N BACKWASH WATER IS DISCHARGED TO SETTLING TANKS & NEUTRALIZIN G TANKS INCLUDED IN THE ACTIVE TREATMENT PLANT

( t ) TEMPERATURE INDICATOR

@ C O N D U C TIV IT Y CELL

Q IN TEG R A TIN G FLOW METER

( p) PRESSURE INDICATOR (N O T SHOW N O N PUMP OUTLETS)

( s r ) SAM PLING C O N N E C T IO N

( h ) H IGH AN D LOW LEVEL CONTRO LS OPERATING PUMPS

FIG. 6. Diagram of pond water treatment plant.


FROM FILTERS TO C A T IO N |

UNIT?I

I

C O 2 IN JE C T IO N SOOIUMTOWER COLUM N

1___

IN ITIA L INSTALLATIO N SHOWN — — —

t TO EFFLUENT TREATMENT PLANT

FIG. 7. Possible additions to pond water treatment plant.


sent foreseen (though space will be left so that, if found necessary, such a plant may be added). Pond activity levels will be controlled by filtration and by occasional purging of pond-water as required.

Table X lists the main parameters of pond-water treatment at Oldbury, Dungeness "B" and Hinkley "B", respectively.

3. 1.2. Pressure-vessel cooling-water treatment plant

The internal steel membrane liner of the concrete pressure vessels is cooled by continuous circulation of water through a system of pipes welded to the liner. The chemical condition of the water is controlled and a treatment plant provided to remove insoluble and soluble corrosion products, these being radioactive owing to neutron activation; Table XI gives basic parameters for the four CEGB pressure-vessel cooling-water treatment plants.

At present it is not considered that regeneration of the ion- exchange columns of these plants will be practicable. The active liquors from the systems will consist of filter and resin-bed back- washings, and occasional drainage for maintenance purposes.

3. 1.3. Humidrier liquorsDrier beds are used to control the moisture levels in the reactor

coolant gas, the drier material usually being activated in alumina in pellet form at magnox stations and special silica gel in AGRs. The liquors collected in the process of periodic regeneration of the drier beds, constitute a significant proportion of the radioactivity in liquid- waste arisings, owing to their relatively high tritium content (about 1 Ci/1). Magnox reactors normally operate with circuit-gas moisture levels of only a few parts per million and in these conditions the volume of drier liquor produced is approximately one litre per day. On occasions of boiler leakage, however, the moisture co llection rate may exceed 10 litres/d .

In the AGRs methane injection will be used to inhibit graphite oxidation; the reaction of methane with carbon dioxide produces water, and the drier bed collection rates may amount to several hundreds of gallons per day.

3.1.4. Other sources ^

The following sources give rise to liquid wastes of very low radioactivity content:

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(i) Laundering of contaminated clothing(ii) Decontamination of plant and equipment(iii) Wash-down of fuel-element transport flask(iv) Washings from personnel changing rooms(v) Washings from floor decontamination

3.2. Collection, treatment and discharge

3.2. 1. Wastes other than humidrier liquors

In general the levels of radioactivity in liquid-waste arisings at magnox stations are such that treatment for removal of soluble activity is not required. It is expected that this will also be the case at the AGR stations. Therefore, waste treatment in general consists of neutralization (if required) and filtration prior to discharge.

At Bradwell and Trawsfynydd stations, however, special limitations are imposed by the characteristics of the Blackwater Estuary and Trawsfynydd Lake, respectively, to which the wastes are discharged; an ion-exchange plant is therefore included in the waste- treatment systems of these stations.

An active-effluent treatment plant is provided at each station to which all active or potentially active liquid wastes, with the exception of humidrier liquors, are passed for treatment prior to disposal by discharge. The wastes are drained or pumped to the plant via installed pipework. The active-effluent treatment plant is normally located in a section of the active ancillaries building, adjacent to, though separate from,the cooling-pond-water treatment plant.

Figure 8 shows a typical treatment plant (in schematic). The incoming wastes are received in primary monitoring and delay tanks; regenerant liquors from ion-exchange columns are collected in a separate tank for neutralization. Filter backwash liquors are also received in separate "settling" tanks from which the settled sludges can be transferred to solid-waste stores and the supernatant liquors pumped off for treatment.

(N.B. In connection with Fig. 8 it should be noted that segregation of soapy and non-soapy wastes (to allow for pre-filtration flocculation of the former) has not been found to be necessary and is not proposed, for example, at more recent stations.)

Treatment consists of filtration by sand pressure or ceramic/cloth filters or both types in series. Table XII summarizes the parameters of the effluent treatment plants at Oldbury and Dungeness "B'.'

133


TABLE X. POND-WATER TREATMENT DATA FOR OLDBURY, DUNGENESS "B " AND HINKLEY "B"

Oldbury Dungeness "B" Hinkley "B"

Volume of pond (gal)

560 000 140 000 202 000

Pond-treatment: a Filtration Ceramic/cloth filters

Porous stainless-steel filters (with pre-coat if required)

Fine ceramic/cloth filters

Ion exchangeCation then through scrubbing tower to mixed bed

NIL NIL

Treatment throughput (gal/min)

17 Reception bay: 33 Main pond: 33 Total: 66

68

Daily treatment as percentage of pond volume

4.4%Reception bay 240% Main pond 40% 48%

a Treatment for AGR ponds may include purging at 1% or 2%/day as required.


T A B L E X I. D A T A ON P R E SSU R E -V E SSE L C O O L IN G -W A T E R T R E A T M E N T P L A N T S (D ata p e r r e a c to r )

Station Oldbury Wylfa Dungeness "B" Hinkley ” B"

Treatment rate(as percentage o f mainflow through coolers

15 3 3 -

Treatment rate (g a l/m in ) 70 133 17 a Not yet

specified

T ype o f filter C eram ic (70 ga l/m in ) Centrifuges P re-coat filters C eram ic filters

T ype o f ion exchange Mixed bed (17 ga l/m in ) Mixed bed M ixed bed M ixed bed

a Pressure-vessel coo lin g system divided into two parts, v iz . active and n on -active ; above data refers to a ctiv e part.

utn


FILTERBACKWASH

FIG.8. Diagrammatic arrangement o f effluent treatment plant.


TABLE XII. DATA ON ACTIVE-EFFLUENT TREATMENT PLANTS AT OLDBURY, WYLFA, DUNGENESS "B " AND HINKLEY "B "

Oldbury W ylfa Dungeness "B" Hinkley "B"

Throughput (g a l / m in)

34 100 17 34

Treatment Sand-pressure filters backed up by cera m ic -c lo th filters

Final monitoring and delay tanks

Number 2 2 4 -

Capacity (ga l) 22000 each 30000 each 8000 each -

Final discharge rate (ga l/m in ) 133 250 67 200

In the case of B radw ell, the additional plant (see para . 1) co n sists o f a non-regenerable ion-exchange unit (cationic) o r "sequ estering" unit fo r treatment of ion -exchange-colum n regenerant liquors from the pond-w ater treatm ent plant. At Traw sfynydd, a s im ila r unit is p rov ided , together with cation and m ixed -bed (anion and cation) units fo r treatm ent, as required, of other radioactive liquid w astes. (See F ig . 9 .)

F rom the treatm ent units the w astes p ass to fin a l-m on itorin g and delay tanks. Subject to the resu lts of m onitoring they are then recircu lated for further treatment or discharged to the main cooling- water discharge system and thence to sea, estuary or lake.

FILTERS

SODIUMCO LUM N

-oMIXED BED UN IT

~ ot • ’

RECIRCULATING LINE

FIG. 9. Possible additional plant required where severe limitations are imposed on activity discharged.

137


3.2. 2. Humidrier liquors

Owing to their sp ecia l nature (high tritium and low particulate content) and mode of arising these liquids are not treated in the active-e fflu ent treatm ent plant. At m ost m agnox stations they are co lle cted (from the d r ie r catchpots) in sm all con ta in ers, sam pled and stored to await measurement of their radioactivity content. They a re then passed into the fin a l-m on itorin g and delay tanks and d is ch arged with the other active effluents.

At the AGR stations, the volum e arisings of hum idrier liquors w ill be com paratively la rg e , and sp ec ia l hold -up tanks a re th e re fo r e installed to which the d r ie r liq u ors a re drained by insta lled pipew ork, break tanks, e tc . A fter m onitoring they a re d ischarged to the coo lin g -w a ter system by an independent p ipeline.

A ctive -e fflu en t treatm ent-plant requ irem ents and equipment

£a)_ Plant_ capacity

The capacity of the active-effluent treatment plant must be adequate to deal (in the 8 -h -d shift) with the estim ated maxim um daily a r is in g s , typ ica l estim ates o f which are sum m arized in Table IX.

C ollection tanks, including fin a l-m on itor in g and delay tanks, a re o f not le s s than 24-h capacity , based on estim ated m axim um a ris in g s , and are n orm ally duplicated. The d isch arge pum ps a s socia ted with fin a l-m on itor in g and delay tanks a re duplicated and a re n orm ally capable o f d isch arg in g the contents of one tank in a p er iod o f tw o h ours , thus enabling favou rab le effluent d isp ers ion conditions, e .g . a particu lar state of tide, to be used to fu ll advantage.

_[b)_ Other requirem ents

T hese m ay be sum m arized as fo llow s:(i) Segregation in prim ary co llection tanks, o f liquids which

req u ire sp e c ia l pre-treatm en t, e . g . neutralization (ion - exchange regenerant liquors) or coarse filtration (laundry wastes containing clothing fib res ).

(ii) Where liquids may be contaminated with oil, means of r e m oval p r io r to or at reception into the treatm ent plant is req u ired .

138


(iii) F acility fo r maintenance work on plant units without in terferen ce to operation of the plant as a whole.

(iv) P rovision for detection of leakage and oversp ill from tanks and pipew ork and fo r prevention of escape of liquors into the ground.

(v) F a cility fo r stirr in g o r recircu la tion and representative sam pling of a ll tank contents.

(vi) Adequate sh ielding o f plant and loca tion o f con tro ls and instrum ents outside sh ielding.

(vii) F a c il it ie s fo r rem ov a l o f sludges, spent f i lt e r and ion - exchange m aterials and fo r their transfer to active-w aste storage vaults o r (See Section 2. 1 .2) p re -s to ra g e trea tment units.

(viii) Adequate instrum entation fo r plant operation ,i. e . d if fe r entia l p re ssu re m easu rem en t a c r o s s f i lt e r s , pH c e lls , conductivity m eters (for ion -exchange colum ns), liqu id - le v e l 'in d ica to rs , e tc .

(ix) D ischarge arrangem ents must be such as to ensure good m ixing o f the a ctiv e -w a ste liq u ors with the con d en ser cooling-w ater discharge and also that there is no poss ib ility of contaminating the cooling-w ater system in the station. The arrangem ents must include p rov ision fo r continuous sam pling fro m the d isch a rg e line throughout the p er iod o f d isch a rg e .

4. GASEOUS WASTES

4 .1 . Sources, types and quantities (See Tables XIII and XIV)

The sou rces of gaseous w astes arising in norm al station opera tion are d escribed below .

4 .1 .1 . Shield cooling a ir

At the s tee l p r e s s u r e -v e s s e l re a c to r stations, coo lin g o f the p re s s u r e -v e s s e l ex terior , charge standpipes and the p rim ary co n crete b io log ica l shielding is achieved by induced a ir -flow over these m em bers. The a ir becom es activated owing to irradiation, gaseous nuclides, notably 41A , being produ ced . Particu late activ ity (about

139


T A B L E XIII. SUM M ARY OF GASEOUS W ASTES A T DUNGENESS " A " , W Y L F A AND DUNGENESS " B "(i) D u ngeness " A " - G aseous E ffluent (n orm a l op era tion )

Source Frequency Quantities (per reactor)

Treatment ' E fficiency Discharge point (s)

Height o f discharge point above ground level

(ft)

1 . S h ie ld -coo lin g air (A ir)

Continuous 280 000 ftVnrin Filtration(Resin bonded fiberglas)

98ft at lO^m Stacks on roof o f reactor building

168 .5

2 . Circulator seal o il degassing and purging(C O J

Continuous 5 t /d per reactor

Filtration(Resin bonded fiberglas)

98ft at 10 (jn Stacks on roof o f reactor building

168 .5

3 . V entilation o f reactor areas (air)

Continuous 15 000 f t 3/m in Filtration - absolute a

9 9 .99 ft W all louvres on sides o f reactor buildings

Varies from 12-50

4 . M ain reactor blowdown(CO*)

1 per reactor per year

100 t Filtration (cera m ic candles)

100ft at 6 fim 98$) at 4 jim 95ft at 2 pm

Stack on roof o f reactor building

140

5 . Ancillary reactor systems blowdown (CO*)

D aily Filtration (ceram ic candles)

100ft at 6 jjm 98ft at 4 fim 95ft at 2 nm

Stack on roo f - o f reactor building

140

a Paper with asbestos fibre and corrugated paper spacers.


T A B L E XIII (con t. )( i i ) W ylfa - G a seou s E ffluent (n orm a l op era tion )

Source Frequency Quantities Treatment E fficiency Dischargepoint

Height above ground (ft)

1 . Reactor main blow dow n (CO*)

O nce - 2 yr 230 t(per reactor)

Filtration (sintered bronze)

1009) 2 10 jim Stack above reactor hall roof

195

2 , Reactor ancillary blowdown (C O j)

D aily Approx, 1 1 (per reactor)

Filtration (sintered bronze)

100ft {§) & 10 (jm Stack above reactor hall roof

195

3 . A ctiv e areas ventilation (air)

Continuous 112000 ft^rnin (tota l)

Filtration Absolute f ilte r a on extract - 9 9 ,99 ft

W all louvres on side o f reactor building

75 - 85

4 . Dry fu e l storage c e ll coo lin g -a ir discharges (air)

Continuous 900 lb /m in Unfiltered 190

a Paper with asbestos fibre and corrugated paper spacers.


T A B L E XIII ( c o n t . )( iii) D u ngeness nB n

Source Frequency Quantities Treatment E fficiency Dischargepoint

Height above ground (ft)

1 . Main reactor blowdown (C O z)

O nce - 2 yr Filtration:Sintered stainless- steel candles

lOO^o @ 1 jim 9 8 7 o @ 0 .4 /im

Stack on reactor roof

240

2 . Reactor ancillaries blowdown (C 0 2)

Approx. w eekly N /A Filtration:Sintered stainless- steel candles

100<7o(g>l/im9 8 7 o @ 0 .4 jjm

Stack on reactor roof

240

3. A ctiv e areas ventilation (air)

Continuous 230 000 f t 3/m in Absolute filters 9 9 .5 ft Stack on reactor roof

240


T A B L E X IV . D A T A ON S H IE LD -C O O LIN G -A IR DISCHARGE A T B E R K E L E Y , B R A D W E L L AND HIN KLEY PO IN T "A "

Station Berkeley Bradwell Hinkley Point "A "

Rate o f flow o f air per reactor (ft3/minj

100800 216 000 212 000

Air temperature o f filter

95°F 100°F 158°F

Treatment Filtration: Bonded glass- fibre blanket, o il impregnated

Filtration: Bonded glass- fibre blanket, o il impregnated

As for Bradwell

E fficiency 100% at 10 93. fPjo at 5 (at 65°F)

97% at 2 10 fan at 100 °F

84^o at 2 - 10 jan 92.5^0 at > 10 jjm

(at 158“F)

No. o f discharge points (per reactor)

2 2 1

Height o f discharge point above ground

( « )

175 137 178

10 C i/h ) is a lso p resen t togeth er with sm a ll quantities o f C 0 2 r e leased by leakage o r con tro lled purging from the re a cto r p ressu re c ircu it .

4 . 1 .2 . M ain and a n cilla ry C 0 2 p ressu re c ircu its

Apart from the infrequent d ischarge, say once every two years f o r m aintenance p u rp oses , o f the tota l p r e s s u r e -c ir c u it c a r b o n - d ioxide content, o r la rg e fra ct ion s ( e .g . b o ile r section s ) th ereo f, sm all d isch arges of coolant gas are made on a d a y -to -d ay basis by reason of, e .g .

(i) F u e llin g -m a ch in e and a n c il la ry equ ipm en t o p e ra tio n s(ii) B u rs t -C a rtr id g e D etection (BCD) sy stem op era tion

143


(iii) C ontrolled purging o f re a c to r p re ssu re c ir cu it , e .g . to r e lie v e c ir cu it o v e r p r e s s u r e , c o n tro l le v e ls o f ca rb on monoxide o r to prevent o il ingress to circu it from the gas c ircu la tor sea ls .

(iv) D egassing of c ircu la to r sea l o i l to rem ove C O 2 absorbed from c ircu it .

4. 1 .3 . Contaminated areas ventilation system s

Contaminated o r potentially contaminated areas are maintained at negative p ressu re relative to inactive areas. Ventilation of such areas g ives r is e continuously to la rge volum es of extracted a ir fo r which h igh -efficien cy filtration is required . The arisings vary con siderab ly from station to station and Table XIII includes quantities fo r two m agnox stations. In som e ca se s the a ir is p r e -f i lt e r e d at intake to m inim ize dust loadings on plant and extract fi lte r s . V entilation fans are located downstream of f i lte rs .

(N .B . In general, the extract filte rs on active area ventilation system s at the latest stations w ill be of a com bustib le type so that they can be incinerated when rep laced . W here high activ ities m ay be expected, how ever, as in the case of irradiated fuel-dism antling ce lls in AGRs, the filte rs provided are n on -com bu stib le .)

4. 1 .4 . P re s s u re -v e s s e l cooling system

Deaeration o f the cooling-w ater system s, at concrete p ressu re- v e s s e l stations, to rem ov e gaseou s produ cts of ra d io ly s is , g ives r is e to a continuous so u rce o f lo w -a ctiv ity effluent.

4 .1 .5 . In cin era tor o ff -g a s e s

See Section 2 .3 .4 .

4 .2 . T reatm ent and d isch arge

Treatm ent of rad ioactive gaseous effluents p r io r to their d is posa l by d ischarge to atm osphere, con sists generally o f filtration , the ch oice of fi lte r type and e ffic ien cy being lim ited by the physical and ch em ica l nature o f the waste and by the throughput requ ired . Treatm ent system s fo r the main gaseous w astes at three CEGB sta tions are sum m arized in Table XIII.

144


S h ie ld -co o lin g -a ir (s te e l p r e s s u r e -v e s s e l stations only) d is ch arges a re m ade fro m stacks above the re a c to r building ro o f. T able XIV g ives data on these d isch arges fo r the f ir s t three CEGB m agnox stations.

R ea ctor-coo la n t gas is continuously filte red on a bypass s y s tem in the re a c to r . M a jor d isch a rg es a re fro m stack s above the rea ctor building; pipew ork and stack dim ensions are such as to ensure adequate dilution and d isp ers ion of the effluent to safe lev e ls . Table XV lists the bypass and discharge filter types provided at each station.

TABLE XV. BYPASS AND BLOWDOWN FILTER MEDIA AT CEGB MAGNOX STATIONS

Station Bypass filters Blowdown filters

Berkeley Cyclones Ceram ic candles

Bradwell Cyclones Ceram ic candles

Hinkley Point " A” C eram ic candles Ceram ic candles

Dungeness "A " Cyclones C eram ic candles

Ttawsfynydd Sintered iron Sintered bronze

Sizew ell Sintered iron Sintered bronze

Oldbury Cyclones C eram ic candles

Wylfa Sintered iron Sintered bronze

T yp ica l Filter E fficiencies: C yclones or 100% rem oval at 10 fjmA m ulticyclones 95-98% rem oval at > 2|im

Quantity fibre 100% rem oval at 6 (jmor ceram ic candles 95% rem oval at a 2 (im

Sintered metals Absolute cut o f f at 4 jim99.5% rem oval at £ 2 fjm

145


It m ay be noted that, in the event of rea cto r gas p ressu res b e com ing e x c e s s iv e , autom atic d isch a rge o f the gas would be m ade to atm osphere via safety va lves, the discharge lines from which in corp ora te f i lt e r s . In addition, a ll stations have plant which would be used, in the event o f a ser iou s re le a se o f f is s io n produ cts into the gas c ir cu it , fo r the treatm ent o f r e a c to r coolant p r io r to d is ch arge .

Small day-to-day discharges o f reactor coolant are made either from p ipes above the re a c to r building o r at som e stee l p r e s s u re - v e s s e l stations, are ducted into sh ie ld -co o lin g a ir o r a ct iv e -a re a ventilation extract sy stem s.

The extract air from active-area ventilation is norm ally filtered at high e ffic ien cy by m eans of "abso lu te" f i lte r s , p re -f i lt e r s being em ployed as n ecessa ry .

A C K N O W L E D G E M E N T S

The ph ilosophy and techniques of waste m anagem ent outlined in th is paper have been developed as a resu lt o f c o -o p e ra t io n b e tw een the United K ingdom A tom ic E nergy A uthority , the N uclear C onsortia and the South o f Scotland E lectric ity and Central E le c tr ic ity Generating B oards.

Thanks are expressed to the South of Scotland E lectric ity Board fo r operationa l data kindly prov ided by H unterston P ow er Station; thanks are a lso due to the w riter 's co llea gu es in the CEGB.

R E F E R E N C E S

[1 ] HER MAJESTY'S STATIONERY OFFICE, The Nuclear Installations Act 1965, HMSO, London.[2 ] HER MAJESTY'S STATIONERY OFFICE, The Radioactive Substances Act 1960, HMSO, London.[3 ] PRESTON, A . , "Site evaluations and the discharge o f aqueous radioactive wastes from civ il

nuclear power stations in England and Wales'*, Disposal of Radioactive Wastes into Seas, Oceans and Surface Waters (Proc. Symp. Vienna, 1966), IAEA, Vienna (1966) 725.

146


APPENDIX 4


APPENDIX 4

UNITED STATES PRACTICE IN

MANAGEMENT OF RADIOACTIVE WASTES AT NUCLEAR POWER PLANTS*

MORTON I. GOLDMAN

NUS C orporation 1730 M Street, N .W .

Washington, D .C . 20036 United States o f A m erica

1. INTRODUCTION AND GENERAL DISCUSSION

The m anagem ent o f ra d ioa ctive w astes p rodu ced at nuclear pow er plants has been the subject o f a con sid era b le amount of in terest to pow er-plant design ers and op era tors , to the regu latory authorities and to the general public o f the United States of A m erica . This in terest has expressed itse lf in the extensive research and development p rogram s on m ethods fo r decontam inating radioactive w astes, in regulations lim iting the concentration and total quantity of waste that may be released to the environment, and in extrem ely conservative design and operating p ra ctice s . A s a result of this combination of governm ental and industrial attention, the experience accumulated to date in the nuclear pow er industry has been u n iform ly excellent. In no case have wastes from a nuclear pow er station released either to the a ir o r w ater environm ent approached the lim its im posed by regulations and the resulting exposures to individuals living adjacent to such fa c ilit ies has been a very sm all fraction of the lim its sp ec ified in US regulations.

In January 1968, B lom eke and Harrington published a report [1] on M anagem ent o f R ad ioactive W astes at N uclear P ow er Stations. This report provides an excellent review of the design and operating

* This appendix was prepared under number NUS-453 by NUS Corporation, Washington.

149


experience in waste management at six nuclear power stations. Much of the inform ation contained in this paper is extracted from this document and the w r ite r1 s appreciation is expressed to the authors of that paper. Additional inform ation was obtained from a previous study published in M arch 1960 by the Edison E lectr ic Institute. This paper [2] a lso d escrib ed the system s in use o r proposed fo r use at stations then in operation o r under construction . Acknowledgem ent is also made to the authors o f that report fo r use of m aterial included in this review .

The fo llow in g se ction s trea t the g en era l ch a ra cte r o f w astes produced from boiling-w ater reactors (BWRs) and pressurized-w ater reactors (PWRs) as constructed in the United States of A m erica , and the waste p rocessin g and disposal methods employed in these nuclear pow er stations. S pecific exam ples are given o f system s in op e ra tion at the present tim e la rge ly from the docum ent by B lom eke and H arrington, and of designs at p resen t in operation o r under c o n struction which vary from those used in the past. Operating experience with these w aste-processin g system s is also described and cost e stim ates are provided where these are available. A dditionally, som e indications are given of potential problem s in the waste-management area particu larly as they m ay affect the p ro ce ss design o r location o f nuclear pow er stations.

2. BOILING-W ATER REACTORS

2 .1 . Introduction and gen era l d iscu ssion

In a boilin g -w ater rea ctor the light-w ater coolant is circu lated usually by fo r c e d con vection up through the re a c to r c o r e w here it b o ils under high p re ssu re ; the steam p a sses to a turbine w here it condenses and is returned to the re a c to r c o r e . An a ir e je c to r r e m oves from the turbine con denser a ir and non -con den sib les which are d isch arged to the plant stack through an o ff-g a s system . The condensate is returned to the rea cto r co re through a condensate filter-d em in era lizer system to rem ove corros ion products that o r ig inate in the turbine and condenser as w ell as to protect the reactor against con d en ser tube leaks o r m ake-up w ater im p u ritie s . The water in the rea cto r v e s s e l is recircu la ted to rem ove the heat con tained in the fuel, and a portion o f the rec ircu la tion flow is passed through a re a c to r c lean -u p system to m aintain high re a c to r w ater

150


purity by the rem ov a l o f c o r r o s io n p rod u cts and f is s io n p rod u cts from the rea ctor w ater. In present large BWR designs the rea ctor coolant clean -u p system a lso em ploys a f i lte r -d e m in e ra liz e r unit.

Under norm al operation, corros ion products and activated gases constitute the m ain so u rce o f ra d ioa ctive w astes fr o m the B W R s. The corros ion products w ill vary depending upon the m ateria ls used in construction o f the prim ary system , turbine and feedw ater heaters, but w ill usually contain rad ioactive isotop es o f coba lt, m anganese, chrom ium and iron . In the absence o f leaking fu e l rod s , 13N, 16N and 190 are the p r im a ry activated gases stripped from the rea cto r co re by the steam flow and carried via the condenser a ir e jector and a delay line to the stack. 13N is by fa r the predominant activity with a calculated re lea se rate of approxim ately 400 juC i/s fo r a nom inal 1000 MW(e) station. In addition to the co rros ion products and a ctivation products d escribed , it is norm ally expected that sm all amounts o f "tram p" uranium contam ination on the fu e l-rod cladding w ill r e sult in the p re se n ce o f m in or quantities o f f is s io n p rod u cts in the re a c to r w ater and the a ir e je c to r d isch arge fro m the turbine c o n d en ser. The turbine gland sea l is a lso norm ally fed with prim ary steam and the gland seal condenser is exhausted into the plant stack v ia holdup p iping. When fu e l-r o d d e fects o c cu r , f is s io n produ cts d iffuse into the re a cto r w ater. The fis s ion ga ses , krypton and xenon, are stripped from the water by the steam and trave l via the turbine gland sea l and the turbine condenser a ir e je ctor to the plant stack. N on-gaseous fis s io n products tend to rem ain in the coolant water where their concentration is controlled by the reactor coolant c lean -u p sy stem . A ny c a r r y -o v e r o f such m a ter ia ls through the turbine and condenser is rem oved in the fu ll flow condensate f i lte r - dem inera lizers b e fore return to the rea ctor .

In general, waste system s are designed to handle the maximum expected activ ity lev e ls without exceed in g app licab le governm ent regulations fo r waste d isch arge . The design ob jective fo r BWR w aste-m anagem ent system s is to maintain average radiation doses beyond the plant boundary from routine plant operation to 1% or less o f the p erm iss ib le exp osu res. In general this design ob jective has been ach ieved in the b o ilin g -w a ter plants operating at the p resen t tim e.

2 .2 . S p ecific BWR exp erien ce

The follow ing sections on D resden 1, Big Rock Point and Humboldt Bay have been taken from B lom eke and Harrington [ l ] . Some additional operating data are provided where available.

151


2_._2_ 1_ _P£.esden_Unit_l

2. 2. 1. 1. General

The D resden N uclear P ow er Station, Unit N o. 1, owned by Com m onwealth E dison Company, is located near the confluence of the Kankakee and D es P le in es R iv ers , about 50 m iles southwest of Chicago. The rea ctor is a General E lectr ic Company "d u a l-cy c le " , boiling-w ater reactor, designed to operate at 700 MW(th) and 200 MW(e) net [3 ,4 ] . The plant fir s t produced nu clear-d erived steam on 15 A pril 1960, and was p laced in com m erc ia l s e rv ice on 1 August 1960 [5] .

T A B L E I. P L A N T C H A R A C TE R ISTIC S TH A T A F F E C T R A D IO A C T IV E W A STE G E N E RA TIO N A T DRESDEN

Coolant

Coolant cy c le

Type o f fuel

M axim um fuel failures

Control rod materials

Other m aterial exposed to primary coolant

Coolant purification

Light water

BWR dual cy c le

U O 2 (T ype I: Z irca loy clad ; II: stainless- steel c lad ; III: Z irca loy clad)

~ 5% o f core assemblies a

B4C in stainless steel

Austenitic stainless steel (primary equipment)

Z irca loy -2 (fuel structure)S tellite (va lve seats)C r-M o steel (primary steam pipe)M onel and cop p er-n ick e l (feedwater heaters) Admiralty m etal (condenser tubes)

Bypass dem ineralization (two 2 0 0 -ga l/m in m ixed -bed units, non-regenerative)

F u ll-flow dem ineralization (three ISSO-gal/rmn m ixed -bed units, regenerative)

Additives to aqueous systems Primary circuit Secondary circuit C losed cooling -w ater system

None (sodium pentaborate backup)NoneNa.CrO.

a Replaced original 2%B stainless steel.

152


This unit has produced about 7. 5 m illion MWh o f e le c t r ic ity up to 31 D ecem ber 1967, with a capacity fa ctor of about 65% and at power levels up to 675 MW(th), 210 MW(e).

C h aracteristics of the plant that affect the nature of the ra d io active w astes a re su m m arized in T ab le I. The f ir s t c o r e (type I fuel) consisted of 1. 5% -enriched U 02 pellets clad with 30 -m il-th ick Z irca loy -2 . From 7 Novem ber 1962, to 7 March 1963, the plant was shut down fo r p artia l re fu e llin g and in sp ection . O f a tota l o f 464 a ssem blies, ten w ere found to be defective. One hundred and ninety irrad ia ted a ssem b lie s w ere rem oved fro m the co r e and w ere r e p laced with 84 new type I and 106 new type II a ssem b lie s . Type II fuel was 2. 5% -enriched U 02 clad with 1 9 -m il-th ick stain less steel. The second p artia l re fu ellin g o ccu rre d during the p eriod 12 A p r il to 14 June 1964, when 97 irradiated types of I and II assem blies were rep laced by an equal num ber of type III assem blies (1. 83% -enriched Z ir c a lo y -c la d U 0 2). On this o cca s io n , 11 " le a k e r s " w ere found. The third partia l refuelling occu rred during the period 28 M arch to16 May 1965, when 200 types I and II assem blies w ere replaced with type III fuel containing erbium and gadolinium as burnable poison s. During this shutdown, 23 leak ers w ere found.

F igure l1 is a sim p lified flow sheet showing the sou rces , trea tm ent, and d isp osition o f ra d ioa ctiv e w astes at the D resden plant, as w ell as the design estim ates of the volum es and activ ities of the p rin cip a l strea m s. The w aste-p lant p r o c e s s e s and equipm ent are sum m arized in T able II.

The Dresden Nuclear Pow er Station operates under AEC License D PR -2 , and radioactive-w aste d ischarges are lim ited by regulations set forth in T itle 10, Code of F ed era l Regulations, Part 20, and by those o f the Illin o is Departm ent of P ublic Health, G aseous-w aste d isch arges are re s tr ic ted by the AEC licen ce to an annual average rate of 700 000 n d / s fo r noble ga ses . T here is no lim it stipulated in the licen ce fo r re lease of iodine and particulates. State of Illinois Department o f Public Health regulations requ ire that liquid d ischarges from the plant do not exceed 10‘ 7 p C i /c m 3, taking into account the background activ ity in the r iv e r w ater used fo r dilution.

The perform an ce o f the rad ioactive -w aste system at D resden has been quite sa tis fa ctory . R evisions in the orig ina l liqu id-w aste system have perm itted g rea ter than 80% reu se o f the w aste w ater; operating costs have been le s s than 10% o f the total operating costs

1 A ll figures are at the end o f this appendix.

153


T A B L E II. W A STE P L A N T PRO CESSES AND E QU IPM EN T A T DRESDEN UNIT NU M BER 1

Gaseous wastes:

Holdup tim e for air e jector gases

Holdup tim e for gland-seal condenser and gases

Air flow in stack Air discharge ve locity Stack height Air ejector gas filter

~ 2 0 min

2 min

44 000 ft3 /m in 50 ft/s .300 ft "Absolute"

Liquid wastes:

FiltersDem ineralizer

Three, 2 0 0 -g a l/m in pressure precoat type One, 2 0 0 -g a l/m in m ixed-bed type, no

regeneration

Concentrator One, 2 .5 -g a l /m in stainless-steel type, natural circulation

Centrifuge One, 3 0 - in . - i .d . by 2 5 -in .-d e e p basketWaste storage capacity a 100 000 galTotal tank capacity a 480 000 galWater available for diluting 167 000 ga l/m in

waste effluent

Solid wastes:

BalerConcrete vault storage Dry-waste shelter Storage building

16 X 24 X 12 in.. bales T w o, 1200-ft3 vaults 2500 ft3 2000 ft3

a Including spent resin and filter sludge storage capacity .

fo r the station. A m a jor sh ortcom ing , the undersized evaporator, w ill be solved by installing a 2 6 -ga l/m in m achine fo r Units 2 and 3, which are currently under construction. It is expected that discharges o f radioactive waste to the r iv e r w ill be substantially reduced in the future by using this equipment to concentrate c o r r o s iv e w astes fo r e con om ica l packaging and o ff -s ite d isposa l.

154


2. 2. 1.2. Solid wastes

2 .2 .1 .2 .1 . Sources

R adioactive solid w astes w ere estim ated [2] to com p rise up to 900 ft3 p er year of spent resin containing as m uch as 4000 Ci; and up to 400 ft3 per year (after baling) of rags, paper, p lastics, filters and assorted rubbish, sm all too ls , equipment, etc . The resin s d e r iv e from the re a cto r coolant clean -u p system (two 200 g a l/m in d e m inera lizers) and the waste clean-up dem ineralizer (one 200 ga l/m in unit) which are not regenerated , and infrequently from the condensate dem ineralizers (three 1550 ga l/m in units) which are regenerated. Each batch of rea ctor coolant clean-up resin was estim ated to contain about 1000 C i.

Spent condensate resin was estim ated to contain about 10 Ci per batch. Spent resins are stored underground in a 50 000 gal stainless- stee l tank. D ry solid com p ressib le w astes are baled, p laced in sea led card board b ox es and stored in a f ir e p r o o f 1200 f t 3 u n d erground con cre te p it.

2. 2 .1 .2 .2 . E xp erien ce

The volum es o f spent resin s produced annually at D resden and cu rren tly in storage there are given in T able III. The la rgest volum e of spent resins originated from the waste dem ineralizer, which p ro cesses about 2 X 106 gal of waste per resin batch, and is not regenerated. The clean-up d em inera lizer res in s are a lso not regenerated and have about a tw o-year life . The condensate dem ineralizer resins are regenerated at about tw o-w eek intervals and have a serv ice life o f about two years. When the volume of spent resins approaches the existing storage capacity , the res in s w ill be pumped into con crete casks and shipped o ff -s ite fo r burial by a licensed con tractor. The expected cost of this operation is about $50 000, an average of about $6000 per year.

Some 2960 f t 3 of diatom aceous earth m ateria l backwashed from the filte rs had accum ulated in the concentrated waste tank at the end of 1967.

A total of 32 477 ft3 of radioactive solid wastes has been shipped o ff -s ite fo r bu ria l (Table IV). The average co s t o f this se rv ice has been about $3000 p er y ea r.

155


T A B L E III. SPE N T RESINS IN STO R AG E A T DRESDEN N U C LE A R PO W E R STATIO N a

PeriodRadwaste

(ft3)Cleanup

( ft 3)Condensate

(ft3)T otal( f t 3)

A ccum ulativetotal(ft3)

1959 0 0 105 105 105

1960 400 160 120 680 785

1961 627 .5 240 82.5 950 1735

1962 640 80 105 825 2560

1963 700 130 270 b 830 3390

1964 400 80 - 480 3870

1965 480 - - 485 965 4835

1966 560 160 - 720 5555 c

1967 480 - - - - 480 1053 d

From Ref. [6 ] .b This batch o f condensate dem ineralizer resins (270 ft3) was shipped o ff site for

permanent burial.c 500 ft3 o f these resins are stored in concrete shipping casks in the radwaste area in

preparation for shipping. d 4982 ft3 o f spent resins were shipped o ff site for permanent land burial.

It is of in terest to note that although the estim ate o f spent resin storage volum e requ irem ents has been reasonably good (averaging 800 ft3 per year), the estim ate of 400 ft3 per year o f m iscellaneous so lid waste has been exceeded , on an average, by a fa ctor of about eight.

2. 2. 1 .3 . Liquid w astes

2 .2 .1 .3 .1 . Sources

As designed at present, the liquid-w aste system perm its seg re gation of waste stream s a ccord in g to th eir ch a ra c te r is t ic s and the type of treatment required. The reactor enclosure piping and equip-

156


T A B L E IV . R A D IO A C TIV E SOLID W A ST E SHIPM ENTS TO PE R M A N E N T LAN D B U R IA L 3

PeriodEstimated activity

(C i)Total(ft3)

A ccum ulative total (ft3)

1960 0 .1 1147 b 1 147

1961 63 5112 c 6 259

1962 5 6 1 d 3312 9 571

1963 1 .1 5970 15 541

1964 0 .4 ' 1918 17 459

1965 41 e 3524 20 983

1966 0 .6 1994 22 977

1967 1 .3 9500 32 477

a From Ref. [6 ] ,b Includes 300 ft3 o f gravel that was rem oved after a radioactive liquid spill. c This batch includes waste accum ulated from revisions and improvements in the

plant, including con trol-rod drive m odifications, radwaste equipment and piping m odifications, and boron poison system changes,

d Disposable steel-and -concrete cask containing 80 control rod blades, 80 in -core cham bers, and m iscellaneous scrap from fuel elem ents accounts for 560 C i o f activ ity and 500 ft3 o f waste.

e Disposable steel-an d -con crete cask containing 12 in -core cham bers, m iscellaneous scrap, sludge, sweepings, and an insignificant unknown quantity o f UO2 pellet m aterial accounts for 40 C i o f activity and 194 ft3 o f waste.

ment drains contain high-purity (low-conductivity) water, which first drains into a 5000-gal ca rb o n -s te e l tank, is then passed through a fi lte r and "w aste d em in era lize r" to either o f two 25 0 00 -ga l holdup tanks, and is finally returned to storage fo r re -u se in the plant. The res in s from this dem in era lizer are not regen erated . Other "c lea n w a stes" p r o c e s s e d by this sam e line o f equipm ent include e x ce ss w ater fr o m startup and w ater fr o m p r im a ry sy s te m b low d ow n s.

157


158

T A B L E V . R A D IO A C T IV E LIQUID W ASTES DISCH ARGED F R O M D RESD EN N U CLE A R P O W E R ST A T IO N 3

High purity M oderate purity Corrosive (neutralized) Laundry

PeriodThousands o f gallons

mCJ 10 4 p C i/litreThousands o f gallons

m C i 10s p C i/ litreThousands o f gallons m Ci 10s p C i/litre

Thousands o f gallons

m Ci 104 p C i/litre

1959 250 <1 <1 339 <1 <1 26 <1 <1 31 <1 <1

1960 1069 120 3 .0 1855 610 0 .9 473 30 0 .2 169 10 1 .2

1961 1088 130 3 .1 1296 810 1 .7 114 70 1 .6 292 30 3 .2

1962 186 10 1 .2 1108 1080 2 .6 506 1270 6 .6 120 10 3 .1

1963 194 40 5 .8 790 1740 5 .8 576 980 4 .5 184 10 1 .6

1964 76 20 5 .2 534 870 4 .3 377 2630 18 .4 87 10 2 .7

1965 0 0 0 41 1590 100.0 263 6630 67 .0 160 40 7 .0

1966 11 <1 <1 87 1089 33 .1 580 10 139 4 6 .2 119 80 17 .9

1967 56 2 1 .0 127 559 11.5 522 3 714 18 .8 294 24 2 .2

Total1 /1 0 /5 0

to

3 1 /1 2 /6 7

2930 322 2 .9 6177 8338 3 .6 3437 25 463 1 9 .6 1456 212 3 .8

a Waste volumes and activ ities measured before dilution in discharge canal and do not include an estimated 5 to 10 C i o f 3H per year [6 ],


The s o -c a l le d "d ir ty w a stes" contain, on the average , h igher concentrations o r ionic im purities than can be rem oved econom ically by dem inera liza tion . T hey con s is t o f the re a c to r en clo su re f lo o r drains, turbine building f lo o r and equipment drains, resin slu icing w ater, eva p ora tor con densates and cen trifu ge o v e r flo w s . T h ese w astes are co lle c ted in a 25 000-g a l c o lle c to r tank and are passed through a precoat-type filter to a 25 000-gal holdup tank. A fter being analysed, they are pumped at a con tro lled rate into the con denser coo lin g-w ater d ischarge canal and re leased to the r iv e r . How ever, if their conductivities are le s s than 50 m icro m h o /cm , they m ay be rou ted fro m the waste f i lte r to the d em in era lize r and returned to condensate storage fo r r e -u s e in the plant.

The water from the secondary steam generator blowdown, is c o l lected in two 5000-ga l tanks. F ollow ing analysis, it m ay either be discharged to the canal or routed to the waste co llector tank and p ro cessed fo r re -u se in the plant.

The "ch em ica l w astes" consist of laboratory wastes, shop floor dra ins, decontam ination so lu tions, and condensate d em in era lizer regenerate solutions. They are s im ila r in that a ll contain con cen trations of ion ic im purities too high fo r e con om ica l rem ova l by ion exchange, and are norm ally collected in a neutralizer tank, adjusted to a pH of 7 to 10 with caustic, and re leased to the discharge canal. If their activ ity is too high fo r d ischarge, these w astes can be con centrated in the evaporator and stored in the 50 000-gal concentrated- waste storage tank.

Laundry wastes are collected in a 1000-ga l tank, are period ica lly sam pled, and are d ischarged to the r iv er a fter being diluted. If their activity levels are high, these wastes may be routed to the neutra lizer tank fo r subsequent evaporation.

The three p ressu re-precoat filters are back-washed periodically into a 1150-g a l sludge r e c e iv e r tank; the sludge is then fed to a centrifuge. On centrifugation the clarified liquid is sent to the waste co lle c to r tank, and the sludge is ploughed out of the centrifuge bowl and dropped into the concentrated-w aste storage tank.

F uel canal w ater is c la r ified by filtration and m ay be returned to the p oo l o r sent to a 25O0OO-gal w aste-storage tank, from which it m ay e ith er be p ro ce sse d fo r r e -u s e o r d isch arged to the r iv e r .

Design estim ates of volum es and activities fo r the several waste stream s a re sum m arized in F ig . 1. The liqu id -w aste system has been m odified by changes in piping and by the installation o f an additional filte r and two centrifuges [7],

1 5 9


T A B L E VI. R A D IO A C TIV E LIQUID W A STE S PRO CESSED FOR R E -U S E A T DRESDEN N U C LE A R PO W E R S T A T IO N 3

T o dem ineralizer T o re-use

PeriodThousands o f gallons

m C i 106 p C i/litre Thousands o f gallons

m Ci 104 p C i/litre

1959 1 665 < 1 <1 1 665 <1 <1

1960 7 728 152 840 5 .2 7 235 390 1 .4

1961 7 764 379 910 12.9 7 753 3180 10 .8

1962 12 640 430 980 9 .0 12 600 1620 3 .4

1963 9 900 545 140 14.5 9 740 540 1 .5

1964 10 020 174 460 4 .6 10 000 190 5 .1

1965 8 260 659 320 21 .0 8 140 4620 15.0

1966 11 590 794 113 18.1 11 470 7046 16.2

1967 21 600 657 100 8 .0 21 480 2473 3 .0

T otal

1 /1 0 /5 9 to 3 1 /1 2 /6 7 91 167 3 793 863 11.0 90 083 20 059 5 .9

3 From Ref. [ 6 ] ,

2 .2 . 1 .3 .2 . E xperience

Liquid w astes, b e fo re dilution and d isch arge , a re given in Table V [6]. The "h igh -purity" wastes are defined as those having a d isso lved -so lid s content of less than 25 ppm, while the "m oderate- purity" wastes range in tota l-so lids content from 25 to 500 ppm. The " co r ro s iv e " wastes are the condensate dem ineralizer regenerant s o lutions, labora tory drains and decontam ination solu tions, and are neutralized to a pH of about 8 b e fo re d isch arge . The "laundry" wastes originate in the plant laundry. These wastes are discharged in batches to the Illin o is R iv er , fo llow in g dilution with con d en ser

160


cooling water to concentrations in the range 10 to 50 p C i/litre (10‘ 8 to 5 X 1 0 '8 (iC i/cm 3 ). During the e igh t-year p er iod rep orted h ere , an annual average o f 1 750 000 gal o f liquid wastes containing 4.3 Ci of rad ioactiv ity w ere d ischarged . The average annual contribution to the r iv e r of radioisotopes ranged from 3 to 38% of the unidentified m ixture lim it. Isotop ic analyses o f selected batches of D resden w aste c a r r ie d out o v e r s e v e ra l y e a rs have ind icated that the d is charge lim it fo r these w astes was about 1 .5 X 1 0 '5 /u C i/cm 3 [8], o r about 150 tim es le s s stringent than the lim it actually used fo r co n tro l. Based on the concentration o f tritium reported in the prim ary w ater (Table XXXIII), it is estim ated that 5 to 10 Ci o f that isotope have been d isch arged p er y ea r in "h ig h -p u rity " w astes, o r about 0 .001% o f the a llow able lim it .

During this same period, 90 083 000 gal of water were recovered fo r r e -u s e in the plant from a total o f 91 167 000 gal p ro ce sse d (Table VI). A ll water p rocessed fo r re -u se at D resden was treated b y filtra tion and d em in era liza tion . The eva p ora tor has not been u sed .

2. 2. 1 .4 . G aseous w astes

2. 2. 1. 4. 1. Sources

The p rin cip a l sou rce o f activity re le a se from the plant is the main condenser air e je c to r . This rem oves about 35 ft3/m in (STP) a ir, ra d io ly tic decom position products o f w ater, and those fiss ion gases and activation gases that m ay be presen t. T hese gases are passed through 130 ft o f 30 -in . diam . pipe, providing about 20 min for decay of short-lived radiogases, before filtration through a high- efficiency particulate filter to a 300-ft stack.

The turbine gland seal condensate exhaust is provided with about 2 min decay by passing through 130 ft of 20 -in . pipe at 200 f t a/m in (STP) b e fo re introduction into the stack.

2. 2. 1 .4 . 2. E xperience

The average annual re lease rates for noble-gas fission products during reactor operation are presented in Table VII. These releases range up to only 3. 6% o f the licen ce lim it o f 700 000ju C i/s . Instantaneous re le a se rates have not exceeded 80 000 /uC i/s [6]. The in cre a se in activity d ischarged during 1964 and 1965 is attributable to

161


T A B L E VII. A V E R A G E ANNU AL R E L E A SE R A T E S OF N O B L E -G A S FISSION PRO D U CTS F RO M DRESDEN NU CLEAR PO W ER ST A T IO N 3

PeriodGross electricity

generated (10s MWh)

Release rate (|iCi/s)

Percent o f licen ce

lim it b

1960 276 800 (m axim um )

1961 555 3 020 (m axim um )

1962 1250 24 300 (m axim um )

1963 989 3 000 (average) 0 .4

1964 1038 20 400 (average) 2 .9

1965 1018 24 800 (average) 3 .5

1966 1475 24 900 (average) 3 .6

1967 853 11 300 (average) 1 .6

a C om piled from Com m onw ealth Edison Com pany Annual Reports for the Dresden Nuclear Power Station,

k Licensed to release at an annual average rate o f 7 x 1 0 5 (iC i/s .

the re latively large number o f sta in less -s tee l fuel cladding fa ilu res that occurred during this tim e. The activities of halogens and particu lates, originated from entrainment in the condensers, have averaged 0. 002 to 0. 003 nCi/s.

_2._2-_2^__Big R ock Point

2. 2. 2 .1 . General

The B ig R ock Point N uclear P ow er P lant, owned by the C onsu m ers P ow er Com pany, is lo ca ted near the c it ie s o f C harlevoix and Petoskey on the north-east shore o f Lake Michigan [9], The rea cto r , a d ir e c t -c y c le , fo rced -c ircu la tion , bo ilin g-w ater unit d e signed to operate at 240 MW(th), 75 MW(e), by the General E lectric Com pany, was con structed by the B echtel C orporation . It firs t reached cr it ica lity in Septem ber 1962 and achieved its in itial rated

162


power o f 157 MW(th), 50 MW(e), in March 1963. The plant has p ro duced about 1418 000 MWh o f e le c tr ic ity up to 31 D ecem ber 1967, with a capacity fa ctor o f about 46% [10], This fa ctor re fle c ts a 50- w eek outage, from 18 Septem ber 1964, to 4 S eptem ber 1965, r e qu ired fo r re p a irs to the r e a c to r th erm al sh ield [11].

Plant ch aracteristics that affect the nature o f radioactive wastes are sum m arized in Table VIII. The f ir s t -c o r e fuel consisted of type 3 0 4 -sta in less -s tee l-c la d , 3. 2% -enriched UO2 pellets, with Inconel- X sprin gs and Z ir c a lo y -2 channels. D uring an outage beginning 9 April 1966, 24 of a total o f 84 irradiated assem blies were rem oved

T A B L E VIII. P L A N T C H A R A C TE R ISTIC S TH A T A F F E C T R A D IO A C T IV E W A ST E G EN ERATIO N A T BIG R O C K POIN T

Light water

BWR; forced circu lation ; single cy c le

Type I; sta inless-steel-clad UO2

Type II: Z irca loy -c la d U 0 2

~ 15% o f core assemblies

B4C powder in stainless steel

Austenitic stainless-steel primary equipment

Z irca loy (fuel structure)Stellite (va lve seats)C opper-n ickel and Admiralty m etal feedwater

heater tubes Admiralty m eta l condenser tubes

Bypass dem ineralization o f reactor water(one 9 0 -g a l/m in unit, non-regenerative)

Full-flow dem ineralization o f condensate(three 6 0 6 -g a l/m in units, regenerative)

Additives to aqueous systemsReactor water None (sodium pentaborate backup)C losed -circu it water coo lin g K 2C r0 4

system

Coolant

Coolant cy c le

Type o f fuel

Maximum fuel failures

Control rod materials

Other m aterial exposed to primary coolant


163


and rep laced with 16 Z irca lo y -c la d bundles and eight new sta in less- stee l-c lad assem blies. During a second refuelling outage in October 1966, 11 defective fuel bundles w ere found and rep laced with Z ir ca lo y - clad fuel.

The B ig R ock Point Nuclear Plant operates under AEC L icense No. D P R -6 , and rad ioactive waste d isch arges are lim ited by regu lations set forth in T itle 10, Code o f F edera l R egulations, Part 20. The m axim um p e rm iss ib le annual average r e le a s e ra te fo r noble and activation-product gases is 1 C i/s . Instantaneous re lea ses for p eriod s up to one w eek are not to exceed 10 C i /s (Table IX ). The allowable annual average rate of release of halogens and particulates m ay not exceed the p erm iss ib le a ir concentrations fo r unrestricted a reas, as given in T itle 10, Code o f F ederal Regulations, Part 20, m u ltip lied by 1. 2 X 1010 cm 3/s . The p e rm iss ib le con cen tra tion s shall be based on isotop ic analyses p erform ed between 48 and 72 h a fter fi lte r rem ova l, and this lim itation includes an allow ance fo r

T A B L E IX . A L L O W A B L E W A STE DISCHARGE AND IN VE N TO R Y LIM ITS A T BIG R O C K POINT

Gaseous wastes, m axim um permissible stack discharge rate:

Noble and activation-product gases:Annual average, C i/s 1 .0Instantaneous, for periods up to

1 w eek, C i/s 10

Particulates and halogens:Annual average, jiC i/s 3 .6

Liquid wastes:

Maximum concentration o f unidentified isotopeso f plant origin in discharge cana l, j iC i /c m 3 10~7

M axim um waste inventory in system , C i 5000

Solid wastes:

Maxim um waste inventory on -s ite , C i 40 000

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annual sta ck -to -g rou n d dilution. B ased on a p e rm iss ib le con cen tration o f 3 XIO"10juCi/cm 3, the maximum rate for halogens and particulates is taken as 3. 6 /^Ci/s.

In the ca se o f liquid w astes, the g ro ss activ ity o f plant orig in may not exceed , on an annual average, the lim its given in T itle 10, Code o f F edera l R egulations, Part 20.

In addition to these re s tr ic t io n s , the AEC lice n ce lim its the liquid waste inventory to 5000 Ci and the on -s ite inventory o f so lid w astes to 40 000 Ci.

T A B L E X . W A ST E P L A N T PRO CESSES AND EQU IPM EN T A T BIG R OCK POIN T

Gaseous wastes:

Holdup tim e for a ir-e jector gases 30 minHoldup tim e for turbine gland seal 90 s

exhaust Air flow in stack Air discharge velocity Stack heightFilter on a ir -e jector o ff-gas

30 000 ft3/m in 40 ft/s (m inim um ) 240 ftH igh -efficien cy

Liquid wastes:

FiltersDem ineralizer

One 7 5 -ga l/m in cartridge type One 7 5 -g a l/m in , m ix ed -b ed , non-

Concentrator Tank storage capacity a T otal tank capacity a Water available for effluent

regenerative type One 1 -g a l/m in , natural-circulation type 15 000 gal 52 000 gal 50 000 ga l/m in

dilution

Solid wastes:

BalerC oncrete vault storage Dry-waste shelter

C om m ercia l unit 1200 ft3

a Including spent-resin storage capacity .

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Figure 2 is a s im plified flow sheet showing the sou rces , trea tment and disposition of all radioactive wastes, with design estimates o f the volu m es and a ctiv ities o f the p rin cip a l strea m s [12]. The w aste plant p ro ce s s e s and equipm ent are su m m arized in Table X .

No significant alterations to the original waste system have been made. Although the system , as constructed, has permitted reasonab ly e ffic ien t segregation o f the w astes, the op era tors b e liev e that even better perform ance in this respect could be achieved by a m ore carefu l separation o f the proper drains and piping. The evaporator is con s id ered to be too sm a ll to be o f m uch p ra c tica l ben efit; h ow ever, it has not been needed to date.

W aste-m anagem ent p erform a n ce at B ig R ock P oint has been sa tisfa ctory in that no significant operational o r maintenance p rob lem s have occu rred , and d isch arges have always been w ell within the allowable lim its . About 80% o f the liquid wastes co llected have been p rocessed fo r r e -u s e by the reactor .

The approxim ate capital co st [1] fo r all rad ioactive w aste fa c ilit ie s at B ig R ock Point was $450 000, o r about 1. 7% o f the total cap ita l co st o f $ 26 700 000. E stim ated operating and m aintenance costs for a typical year w ere $28 000, or 4. 7% of the total plant op erating and maintenance cost of $590 000, exclusive o f fuel.

2. 2. 2. 2. Solid wastes

2. 2. 2. 2. 1. Sources and handling

Solid w astes are p rim arily spent resin s from the cleanup, con densate and waste dem inera lizers ; filter socks and catridges; con tam inated too ls and equipm ent; and m isce llan eou s sm a ller item s. The re s in s a re s lu iced to a 10 0 0 0 -ga l stora ge tank that w as d e signed fo r five years o f plant operation . F ilter ca rtr id g es , too ls , e tc . , are stored in either am underground con cre te vault o r a dry shelter, whereas com pressib le item s are baled and sim ilarly stored, pending o ff-s ite shipment for permanent disposal. Design estim ates are shown in Fig. 2.

2. 2. 2. 2. 2. E xperience

During O ctober and N ovem ber 1965, six shipments of solid r a d ioactive w astes w ere made to an o ff -s ite burial ground. These sh ipm ents con s is ted o f 794 ft3 o f ba led w aste contain ing 0. 22 C i,

166


three drum s o f dry lo w -le v e l w aste containing le s s than 1 Ci, and 768 ft3 o f spent d em inera lizer resin s containing 378 Ci. Since that tim e, an additional 430 f t 3 o f w aste containing 4600 Ci o f activ ity have been shipped. M ost o f the activity in the later shipm ents was associa ted with used re a cto r con tro l b lades that occup ied a volum e o f about 100 ft3.

2. 2. 2. 3. Liquid w astes

2. 2. 2. 3. 1. Sources and treatm ent

The liquid waste system is designed to handle 70 000 gal o f w aste p er day, and each o f the effluents is re le a se d in batches o f 4500 to 5000 gal at ra tes such that the a ctiv ity in the co n d e n s e r - c ircu la tin g -w ater d ischarge canal does not exceed the p erm iss ib le lim its . The co llection system is designed to segregate wastes into th ree gen era l ca te g o r ie s : (1) "d ir ty -w a s te s " , which include f lo o r drains and other stream s with a high concentration o f dissolved inert so lid s ; (2) "c lea n w astes", which are norm ally re a cto r water with a v ery low concentration o f d isso lved so lid s ; and (3) "ch e m ica l w a s tes", which include acid and b a s ic solu tions from the laundry, laboratory , dem in era lizer regeneration and decontam ination o p e r ations. A fter analysis , the d irty w astes and the ch em ica l w astes a re n orm a lly d isch a rged through a nylon so ck f i lte r to the canal; but, i f n ecessary , they can be either filtered and dem ineralized or concentrated and stored tem porarily on site. The clean wastes are genera lly passed through a ce llu lo se -ca r tr id g e filte r , a m ixed -bed "radw aste d em in era lize r" , and returned to condensate storage fo r reu se . The radw aste d em in era lizer is used a lso to decontam inate the spen t-fuel pool w ater when requ ired .

2. 2. 2. 3. 2. E xperience

Liquid waste discharge h istory is sum m arized in Table XI. An average o f 5. 8 C i/y r o f activity, exclusive o f 3H, has been released to Lake M ichigan via the circu latin g-w ater d ischarge canal. M ore than 90% o f this waste was re leased on the basis o f unidentified is o top es, the rem ainder being re lea sed on the basis o f partia l iden tification, w hereby at least 90% o f the activity was found to be a c o m bination o f 58Co and 65Zn.

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T A B L E X I. R A D IO A C TIV E GASEOUS AND LIQUID W ASTES DISCHARGED A T BIG ROCK P O IN T 3

Gaseous releases Liquid discharges

Period Release rate (jjC i/s)

Average Peak0Volum e^ do3 gal)

A ctiv ity(C i)e

Gross electricity generated (MWh)

Sep. 1962 - May 1963 M ay 1963 - Nov. 1963

<20f 20f 50 f 54f

- 6 0- 6 0 827

0 .43 .1 115300

Nov. 1963 - May 1964 - 6 0 560 3 .5 179900May 1964 - Nov. 1964 - 6 0 541 5 .2 103400Nov. 1964 - May 1965 ~0 - 0 433 1.1 0May 1965 - Nov. 1965 5 0008 19 000 414 0 .8 90 300Nov. 1965 - May 1966 30 0008 77 000 310 2.0 246 200M ay 1966 - Nov. 1966 40 OOOS 80 000 202 4 .8 153 700Nov. 1966 - May 1967 10 0008 17 000 296 5 .1 229 000M ay 1967 - Nov. 1967 11400® Not reported 437 2 .0 267 300

a C om piled from Consumers Power Company monthly and sem i-annual operating reports for Big Rock Point Nuclear Plant, b Permissible continuous discharge (averaged over 12 consecutive months) is 1 C i/s .c Permissible instantaneous rate o f release is 10 C i/s .d V olum e before dilution, based on average batch volum e o f 4700 gal.e Generally, m ore than 90°jo o f this activ ity is due to ®Co and 65Zn . An additional 20 C i o f ^ are estimated to have been

discharged per year, f Principally 13N.g~ Greater than 99°jo noble-gas fission products. Halogens and particulates have averaged less than 1. 2 /iC i/s.


Based on a tritium concentration of 0. 03 /uCi/cm 3 in the prim ary coolan t, it is estim ated that about 20 Ci o f tritiu m have been r e lea sed p er y ea r, which is about 0. 007% o f the a llow able lim it fo r that iso top e .

2. 2. 2. 4. G aseous w astes

2. 2. 2 .4 . 1. S ources and treatm ent

N on-condensables from the main condenser air e jector are r e le a se d to the 240 -ft stack through a h ig h -e ff ic ie n cy f i lt e r a fter a holdup o f 30 m in in an expanded o ff -g a s lin e , w hereas the turbine gland seals are vented to the stack after a delay of about 9 0 s . Ventila tion a ir from the ch em ica l labora tory and counting room is filtered and re leased to the atm osphere through a roo f exhaust vent.

With a 240-ft stack and the annual wind distribution found at the site, it was calcu lated that an annual average stack re lea se rate of 1 C i /s , o r a sh o rt-te rm re le a se rate o f 10 C i /s , would not resu lt in off-p lant doses o f 500 m r e m /y r ; and the o ff-g a s system was d e signed with the expectancy that em issions would be w ell below these p e rm iss ib le em iss ion ra tin gs. The a ir e je c to r system rad iation m on itor near the beginning o f the d ecay system a la rm s i f the em ission rate after 30-m in decay exceeds 1 C i/s , and automatically initiates valve closu re with a 15-m in tim e delay if the em ission rate a fter 30 -m in decay should exceed 10 C i /s . In the event that the valve c losu re actually o ccu rs , within a few minutes plant shutdown would be requ ired due to lo s s o f condenser vacuum .

2. 2. 2. 4. 2. E xperience

A su m m ary o f the r e le a s e s o f ra d io a ct iv e gaseou s w astes is presented in Table XI. A verage gaseous re leases to the end of April 1965 ranged from le s s than 20 to 54 /uC i/s and con sisted m ainly of 13N. Since that date, noble gases from defective fuel elem ents have resu lted in substantially h igher re le a se s 'although they have n ever been m ore than 4% of the licen ce lim it. Following the last refuelling outage in O ctober 1966, the rates of gaseous re lease have decreased substantially. The re le a se o f halogens and particu lates has averaged le s s than 1. 2 /nC i/s.

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2 .2 .3 .1 . G eneral

The Humboldt Bay Pow er Plant, Unit N o.3, owned by the P acific Gas and E lectric Company, is located on Humboldt Bay, approxim ately fou r m ile s south o f E ureka, C a liforn ia [9 ,1 3 ]. The r e a c to r is a d ir e c t -c y c le , n a tu ra l-circu la tion , boilin g -w ater unit, designed by the G eneral E le ctr ic Company fo r an ultim ate net pow er operating level of 230 MW (th), 68 MW (e). The reactor first achieved cr it ica lity in F eb ru a ry 1963 and w as p laced in c o m m e r c ia l opera tion on1 August 1963. This unit has produced 1 335 000 MWh of e lectr ic ity up to 31 D ecem b er 1967, with an ava ilab ility fa c to r o f about 0.8 w hile operating at an in itial rating o f 165 MW (th), 52 MW (e).

The plant ch a ra c te r is t ic s that a ffect the nature o f the r a d io active w astes are sum m arized in Table XII. The first co re of type I fuel consisted of 2.6% -en rich ed UO2 pe llets, clad with 19-m il-th ick type 304 sta in less s tee l. During the fir s t half o f 1965, in creasing fission -product ga s-re lea se rates indicated the presence of defective fuel elem ents. During the first refuelling outage from 20 September to 1 D ecem ber 1965, 57 irrad ia ted type I a ssem b lies (out o fa to ta l c o r e loading o f 188) w ere rep la ced with 43 type II (33 -m il-th ick Z irca lo y -2 cladding) fuel a ssem blies . F orty -e igh t o f the irradiated a ssem b lies w ere found to have one or m ore rods with defective c ladd ing , w hich resu lted fro m s tre s s c o r r o s io n cra ck in g o f the cladding m ateria l. During the second refuelling outage, 21 N ovem ber 1966, to 5 January 1967, additional irra d ia ted type I a sse m b lie s w ere rep laced with Z ir c a lo y -c la d fuel, and the operation was r e su m ed , with about a qu arter o f the c o r e con s is t in g o f s ta in le s s - s te e l-c la d fu el. Beginning on 21 F ebru ary 196?; it was n ecessa ry to red u ce the unit load [ 14] to m aintain the o ff -g a s re le a s e rate below the annual average lim it . F ro m 29 M arch to 5 A p r il 1967, the unit was shut down to rem ov e a ll d e fective and h igh -exp osu re type I a s s e m b lie s . T w en ty -th ree o f the h ig h e st-e x p o su re type I a sse m b lie s w e re re p la ce d w ith the rem ain in g tw en ty -th ree low - exposure type I e lem ents. O f the rem oved a ssem b lie s , nine w ere indicated to have defects , one was suspect and thirteen w ere sound. During a scheduled re fu e llin g outage from 4 to 28 O ctob er 1967, 42 o f the 45 rem ainingtype I fuel a ssem b lies w ere rep laced by type II fu el a ssem b lies [15].

2.2 .3 . Humboldt Bay

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T A B L E X II. P L A N T C H A R A C TE R ISTIC S TH A T A F F E C T R A D IO A C T IV E W A STE G E N E RA TIO N A T H U M BO LD T B A Y

Coolant Light water

Coolant cy c le BWR; natural circu lation ; single cy c le

Type o f fuel Type I: stainless-steel- clad UO 2 T ype II: Z irca loy -c la d U O z

M axim um fuel failures 35a/o o f core assemblies (one or m ore fuel rods per assembly)

Control rod materials B4C in stainless steel and 0 .9 % - B stainless-steel curtains

Other m aterial exposed to primary Austenitic stainless-steel primarycoolant equipment

Z irca loy (fuel structure)Stellite (valve seats)Adm iralty m etal feedwater heater tubes

(replaced with stainless steel 12 Nov, 1966)

S ilicon -bron ze condenser tubesheet and aluminium-brass tubes

Coolant purification Bypass dem ineralization (one 5 0 -g a l/m in unit, non-regenerative)

F ull-flow condensate dem ineralization (three m ix ed -b ed , 6 2 5 -g a l/m in units, regenerative)

Additives to aqueous systems

Reactor water None (sodium pentaboiate control system backup)

C losed -circu it w ater-cooling K 2C r0 4system

Suppression pool N a2Cr04

The Humboldt Bay Plant operates under AEC License No.D PR-7. Liquid radioactive w aste d isch a rges are lim ited by regulations set forth in T itle 10, Code o f F ed era l R egu lations, P art 20, and by those of the State of California North Coastal Regional Water Quality C ontrol B oard , w h ichever a re the m o re r e s tr ic t iv e (T ab le XIII).

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TABLE XIII. ALLOW ABLE W ASTE DISCHARGE ANDINVENTORY LIMITS AT HUMBOLDT BAY

Gaseous wastes, m axim um permissible stack discharge rate a:

Noble and activation gases:

Annual average, fiC i/s 50 000Instantaneous, |iCi/s 500 000

Particulates and halogens:

Annual average, jiC i/s 0 .1 8

Liquid wastes:

Maxim um concentration o f unidentified isotopes o f plant origin in discharge canal*3:

From Feb. 1963 to Apr. 1965

Average for any seven consecutive days, / iC i /c m 3 10"8

From M ay 1965 to Apr. 1966

Average for any seven consecutive days, f iC i/ c m 3 10"7Annual average, f iC i/c m 3 10"8

E ffective May 1966

Average for any seven consecutive days, |iCi/cm 3 1 0 '6Annual average, ( jC i /c m 3 10“7

M axim um waste inventory in system, C i 10 0 0 0 a

Solid wastes:

M axim um total waste inventory on site, C i 50 000 a

3 AEC lice n ce values.b Requirements o f the State o f California North Coastal Regional Water Quality

Control Board.

In the ca se o f gaseous w astes, the AEC licen ce lim its are app lica b le . The annual stack re le a se ra te fo r f is s io n product noblegases is 50 000 n Ci / s , with the peak rate being 500 000 y C i / s . The

172


allowable annual average re lea se rate fo r halogens and particulates shall not exceed the perm issib le air concentrations for unrestricted a reas g iven in T itle 10, C ode o f F ed era l R egu lations, P art 20, m ultip lied by 6 X 10° cm 3/ s . B ased on a p e rm iss ib le co n ce n tra tion o f 3 X 10“ 10/u C i/cm 3, the a llow able rate is 0.18/^ C i /s .

T he m axim um concentration of liquid w astes in the plant d is charge canal, as set by the State Board (see above), was orig inally an average o f 10-8 /u C i/cm 3 fo r any seven con secu tive days. Since the start of operations, it has been revised tw ice to le s s -re s tr ic t iv e lim its on the basis of favourable results with o ff-s ite environm ental m on itorin g . C urrently , the average concentration fo r seven co n secu tiv e days and the annual average con cen tra tion m ust not e x ceed 10- 6 and 10_7/L(Ci/cm3 re s p e c t iv e ly .

In addition to these d ischarge lim its , the AEC licen ce restr icts the liqu id-w aste inventory to 10 000 Ci and the total waste inventory o f solid w astes in on -s ite storage to 50 000 Ci.

F igu re 3 is a sim plified flow sheet showing the so u rce s , trea tm ent and d isp osition o f rad ioactive w astes at Humboldt B ay. The a ctiv ities and volu m es shown a re based on operating exp erien ce . The w aste plant p r o c e s s e s and equipm ent (ra d -w a ste system ) are su m m arized in T ab le X IV .

The only sign ificant a lteration to the sy stem has been the r e placem ent o f an orig ina l vacuum precoat radw aste filte r with additional ca rtr id g e -ty p e units. The latter is p re fe rred because the large volum es o f w aste generated by backwashing are avoided, and because o f m echan ica l p rob lem s associated with the precoat filte r . O perators have suggested that future liqu id -w aste system s can be im proved by installation o f duplicate tanks and piping in a m anner that would perm it better segregation of w astes of d ifferent activ ity and conductivity lev e ls .

The total capital cost o f the w aste-m anagem ent fa c ilit ie s (d e fined as fa c ilit ies that would not be n ecessa ry in a fo s s il fuel plant) w as approx im ately $680 000, o r 2.8% o f the tota l cap ita l co s t fo r the Humboldt Bay unit ($14 200 000). Of the $680 000, approxim ately $144 000 w as used fo r stru ctu res and im p rovem en ts, and $436 000 was used fo r equipm ent. The s o l id - , g a seou s - and liqu id -w aste fa c ilit ie s represen ted about 7%, about 16% and about 77% resp ectiv e ly o f the $680 000. T ota l operating costs fo r labour and m ateria l (e x cluding fuel) fo r the station w ere a pprox im ately $397 000 fo r a typ ica l year; and the cost o f m ateria ls fo r w aste-m anagem ent and so lid -w a s te d isp o sa l s e r v ic e s w as $9 340, o r 2.3% o f the tota l operating co s t .

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T A B L E X IV . W A STE P L A N T PRO CESSES AND E QU IPM EN T A T H U M BO LD T B A Y

Gaseous wastes:

Holdup tim e for air e jector gases

Holdup tim e for gland-seal condenser

Air flow in stack Air discharge velocity Stack height Gas filter on air e jector

o ff-gas Caustic scrubber3

Liquid wastes:

Filters

Dem ineralizer

Concentrator Tank storage capacity*5 T otal tank capacity*3 Water available for diluting

waste effluent

18 min design; approxim ately 40 m in actual at rated load

50 sec

12 000 ft3/m in 100 ft/s 250 ft"Absolute" (99 .97% rem oval o f particulates

> 0 .3 |im d iam .)134 ft3/m air, 2 .5 ga l/m in o f 5% caustic

Three 5 0 -g a l/m in , 5-jim cellu lose- cartridge type

One 5 0 -g a l/m in , m ixed -bed type, non- regenerative

One 1 -g a l/m in , natural-circulation type 20 000 gal 62 000 gal 100 000 ga l/m in

Solid wastes:

C oncrete vault storage 1200 ft3Storage building 2000 ft3

a For refuelling-bu ild ing air during refuelling or in event o f contam ination incident, b Including spent-resin storage capacity .

2 .2 .3 .2 . Solid w astes

2 .2 .3 .2 .1 . S ou rces

The prin cipa l sou rces o f solid w astes are spent d em in era lizer res in s , filte r cartridges and m iscellaneous contaminated m aterials from maintenance and refuelling operations. Spent resins are stored

174


in a 10 000-gal resin storage tank. The filte r cartridges are placed in a shielded 55 -ga l drum . When a drum is filled , it is tran sferred to a 1200 f t 3 cap acity , su b su rfa ce , con cre te stora ge vault. O ther w astes a re co lle c te d in standard IC C -approved fib reb oa rd b o x e s , and stored in either the vault o r a building, pending o ff -s ite sh ip ment and d isposa l by a licensed con tractor.

Two m ixed-bed d em inera lizer system s are provided fo r coolant pu rifica tion : a 50 -g a l/m in "cleanu p d e m in e ra lize r " re c ir cu la te s w ater within the re a c to r , and three 625-g a l/m in units (two operating and one spare) p ro cess the total condensate stream from the turbine con den ser. The cleanup dem in era lizer res in s are not regenerated but when depleted a re s lu iced to a r e s in -s to ra g e tank loca ted in the ra d ioa ctive w aste (radw aste) fa c ility . The condensate d e - m inera lizer res in s , on the other hand, are regenerated with caustic and su lphuric a c id , y ie ld in g an aqueous Na2S04 so lu tion . A fte r a u se fu l life o f two to fou r y e a r s , th is re s in is d isca rd e d , b eca u se o f m ech a n ica l degradation , by s lu ic in g to the re s in s to ra g e tank. Up to 31 D ecem b er 1967, fou r 1 7 .5 -ft3 batches o f cleanup d e m in era lizer resin and one 34-ft3 batch o f condensate dem inera lizer resin had been sent to storage.

2 .2 .3 .2 .2 . E xperience

Since the start o f operations, about 22 0 -ft3 of resin s have been accum ulated in the 1300-ft storage tank; this represents an average o f about 50 ft3/y r . It is planned eventually to package and ship them o ff-s ite fo r d isposal. Three shipments of other types of solid wastes, totalling about 930 C i in a volum e o f 3775 f t 3, have been m ade (T able X V ). A ll but about 3 Ci o f this activity was associa ted with 8.7 f t 3 o f po ison cu rta in s, w hich would n orm ally be held f o r lo n g term decay in the fuel p oo l. The cost o f the lo w -le v e l so lid -w a ste shipm ent con ta iners and the d isp osa l operation has averaged about $ 3 .0 0 /ft3, excluding the sp e c ia l poison curtain shipm ent.

2 .2 .3 .3 . L iquid w astes

2 .2 .3 .3 .1 . S ou rces and treatm ent

The liquid w astes a ris in g from the sou rces indicated in F ig . 3 a re co lle c ted in fiv e tanks. Laundry and hot sh ow er w astes are sam pled and norm ally d ischarged d ire ct ly to the canal; but, i f d e -

175


TABLE X V . SOLID-W ASTE SHIPMENTS FROMHUMBOLDT B A Y 3

ShipmentNo.

Date o f shipment

Description o f wasteV olum e o f

waste (ft3)

A ctiv ity o f waste (C i)

1 3 0 /1 /6 4 1 4 4 -4 .5 - f t 3 boxes o f lo w -le v e l waste

649 0 .8 8

1 -cask (2 0 .5 - in .-d ia m . X 46-in . high cavity) o f irradiated poison curtainsb

8 .7 926.76

2 1 7 /1 2 /65 3 5 0 -4 .5 - f t 3 boxes o f low -le v e l waste

1575 0 .3 5

2 7 -5 5 -g a l drums o f filter cartridges

195 0 .6 8

3 2 1 /3 /6 7 2 9 1 -4 .5 - f t 3 boxes o f lo w -leve l waste

5 -5 5 -gal drums o f filter cartridges

1310 'I

37 J 1 .24

3 From Ref. [1 5 ].b Poison curtains were rem oved to permit m odification o f spent-fuel p oo l. H igh -leve l

irradiated materials o f this nature would ordinarily be held in long-term storage in the pool.

contamination is required, they can be com bined with other stream s in the tu rb in e-bu ild in g drain tank. The w astes a re pum ped fr o m the co llection tanks to one of five w aste-treatm ent plant tanks1 where they are m ixed by re circu la tion pum ps, sam pled, and analysed fo r g ro ss beta and gam m a activ ity .

Minimum treatment consists of filtrationthrough 5-/um, 50-gal/m in cellu lose cartridge filters with d irect discharge to the p lant-cooling- w ater ou tfa ll canal. T h ree f i lte r s a re a va ila b le , and they can be

1 The three "receiver tanks" and two "hold tanks" in F ig.3 are used interchangeably for waste storage before and after treatment.

176


used in either a se r ie s o r a para lle l arrangem ent. E xperience has shown that each filte r can p rocess 3 000 to 20 000 gal of waste before cartridge replacem ent is required . If additional treatment is needed, a 50 -g a l /m in m ix e d -b e d , n on -reg en era tiv e d e m in e ra liz e r and a 1-g a l /m in w aste con cen tra tor are ava ilab le . The con cen tra tor is constructed of carbon s tee l and uses cop p er-n ick e l tubes. The d e pleted ion exchange res in is s lu iced to the res in storage tank, and the con centrator bottom s are stored in either o f two concentrated - w aste s tora g e tanks. A fte r treatm ent, the decontam inated w aste is co lle c te d in one o f the treatm ent system tanks, w h ere it is sam pled and analysed. It may then be recyc led fo r additional trea tm ent, pumped to the reactor-w ater-conden sate storage tank fo r r e u se , o r re lea sed to the plant outfall canal, w h ere it is d iluted by a fa c to r o f 103 to 104 with tu rb in e -con d en ser coo lin g w ater .

2 .2 .3 .3 .2 . E xp erien ce

The average liqu id-w aste processin g rate has steadily declined from 3100 g a l/d in 1964 to le ss than 1000 g a l/d in 1967 because leakage from variou s p ie ces o f plant equipm ent has been redu ced to an absolute m inim um . O ver 97% o f the w aste has requ ired only filtration b e fore d ischarge; the rem ainder has required dem ineralization. The concentrator has been tested and used on severa l occasions but has not been requ ired s in ce a ll p ro ce sse d w aste has been d is charged. It has not been p oss ib le to return any p rocessed waste to the condensate stora ge tank becau se o f contam ination by sm a ll amounts o f residu a l h igh -conductiv ity w ater usually present in the treatm ent system tanks and piping.

Liquid d isch arges exceeded the lim its only once, by a very sm a ll m a rg in , fo r seven con secu tiv e days du rin g the p er iod 15 F ebruary - 15 August 1964; since this tim e, C alifornia has lib e r a lized this particu lar restr iction by a fa ctor of 100. Based on a tritiu m con cen tration o f 0.03 fx C i/cm 3 in the r e a c to r w a ter , it is estim ated that about 20 Ci o f this isotope have been discharged per year. This is about 0.004% of the allowable lim it for tritium release .

2 .2 .3 .4 . Gaseous wastes

2 .2 .3 .4 .1 . Sources and treatment

The principal sources of gaseous wastes are the main condenser a ir e jector and leakage from the turbine gland system . Other sources

177


178 T A B L E X V I. R A D IO A C TIV E GASEOUS A N D LIQUID W A STE DISCHARGES A T H U M BOLDT B A Y a

Gaseous waste releases Liquid waste discharged

N oble and activation gases** Particulates andAverage

concentration

Gross

PeriodAverage

ratePeakrate

halogens, range o f w eekly average

Tota l e lectricitygenerated

(fiC i/s) (fJCi/s) (pC i/s) 0 iC i /c m 3) C ic (MWh)

16 Feb. 1963 - 15 Aug. 1963 50 500 6 X 10'“ - 2 X 10‘ 3 0.16 x 10‘ 8 0.16 41600

16 Aug. 1963 - 15 Feb. 1964 34 440 9 X 10"5 - 6 X 1 0 '’ 0.38 x 1 0* 0.38 171100

16 Feb. 1964 - 15 Aug. 1964 94 650 2 X 10_S - 9 X 10"‘ 0.16 x 10'8 0.16 214400

16 Aug. 1964 -1 5 Feb. 1965 390 1225 l x 10‘ s - 4X10"* 0 .3 8 x 10‘ ® 0.38 182200

16 Feb. 1965 - 15 Aug. 1965 12 200 65 000 8 X 10 s - 3 X 10"* 0 .4 1 x 10‘ * 0.41 175 400

16 Aug. 1965 - 15 Feb. 1966 16 000 85000 3 X 10_s - 3 X 1 0 '1 1.40 x 10'* 1.44 62 500

16 Feb. 1966 - 15 Aug. 1966 5 500 43 500 3 X 10 * - 2 X 10"! 0 .6 5 x 10 '8 0.66 84400

16 Aug. 1966 - 15 Feb. 1967 23 000 50 000 1 X 10'* - 1 X 10"! 1 .7 1 x 1CT8 1.74 123200

16 Feb. 1967 - 15 Aug. 1967 23160 Not reported 4 x lO"3 - I X 10"‘ 0. 38 x 10-8 0 .3 8 133 703

16 Aug. 1967 - 15 Feb. 1968 33 300 Not reported 9 X 10‘ s - 9X 1 0 '! 3 .1 0 x 10’ 8 3.16 213 004

aC om piled from P acific Gas and Electric Company semiannual operating reports for the Humboldt Bay Power Plant.

bSee T ab le XIII for allow able lim its.

°A n estim ated additional 20 C i o f SH have been discharged per year.


are a ir fro m purging the dryw ell and from the refu ellin g building. The dryw ell is purged only when a ccess to the reactor, for refuelling o r other p u rp oses , is req u ired . The design o f the expanded o f f - gas lin es p rov id es an 18-m in delay fo r the a ir e je c to r w aste2 and a 50 -s delay fo r the g lan d-sea l exhaust stream b e fore filtration and re lea se through a 250-ft-h igh stack.

The air e jector gases are monitored by two ion-cham ber gamma - d etectors (one is an operating spare) designed to detect radioactive nob le g a s e s . A re le a se rate o f 0.05 C i /s ca u ses an a la rm to be sounded, and if a rate o f 0.5 C i /s is sustained fo r 10 m in, the o f f - gas isolation valve c lo se s and the rea ctor is shut down. The stack gas is m on itored fo r halogens and particu lates with ch a rco a l and particulate f i lte r s , and fo r gam m a-radiation with two scin tilla tion d e te cto rs (one is an operating sp a re ). C a librations o f the o ff -g a s and s ta ck -g a s m on itoring system s a re based on p e r io d ic iso top ic analyses of "grab" sam ples, using a 256-channel gam m a-spectrom eter (scintilla tion detector) system . Both system s perm it evaluation of n o b le -g a s and g a seou s -a ctiv a tion -p rod u ct re le a s e ra tes o v e r the range 102 to 107 /u C i/s .

2 .2 .3 .4 .2 . E xp erien ceRadioactive waste d ischarges (Table XVI) have been consistently

within the allow able lim its given in Table XIII. B efore 15 F ebruary 1965, the average and peak n ob le -g a s re lea se ra tes w ere co n s is t ently le ss than 1% o f the lim its . During the periods since that tim e, which w ere typified by an increasing number of fuel element failures, the re le a se rate has been in the range of 10 to 50% o f the lim its . It is expected that the re le a se w ill rem ain in this range until a ll the s ta in le s s -s te e l-c la d fuel is rep la ced with Z ir c a lo y -c la d fu e l, and residual contamination from "tram p" uranium in the prim ary system is e lim in ated . M axim um w eek ly average particu la te and halogen re le a se rates a lso in cre a se d , fr o m about 1% to a lm ost 40% o f the average annual lim it.

2 .3 . P resen t BW R design p ra ctice

2 .3 .1 . G eneral

T here has been re la tiv e ly little change in BW R w aste system design p ractice since construction of the ea r lie r stations d escribed

2 Actual delay at full load is approximately 40 min.

179


above. Some o f the changes that a re being introduced into stations in itiated o v e r the past y e a r o r tw o a re s ig n ifica n t, but cannot be evaluated until som e operating experience has been obtained. These changes w ill be d iscussed below .

An exam ple of the waste system design typ ical o f boiling-w ater plants starting operation in 1968-1969 is provided by the Nine M ile Point Plant o f the N iagara-M ohaw k P ow er C om pany [16]. This 500 000-kW station (1779 MW (th), located on the south shore o f Lake O ntario in New Y ork , is to be fuelled with Z ir c a lo y -c la d UO2 and operate as a d ir e c t -c y c le plant. Since construction is not com plete at the tim e o f w riting, neither the operating lice n ce lim itations on w aste d is ch a rg e , nor the plant c o s ts , a re ava ilab le . A s im p lified flow sheet o f the ra d ioa ctive -w a ste system s is presented in F ig .4 , and ca p a cities o f m a jor com ponents are listed in T able XVII.

2 .3 .2 . Solid w astes

Solid w astes w ill include filte r sludge, spent res in , concentrated liquid w aste, contam inated too ls and equipm ent, and m iscellaneou s rubbish fr o m la b o ra to ry , m aintenance and cleanup o p era tion s . Estimated annual accumulation and average activities of these wastes are given in Table XVIII.

F ilter sludge and spent resin w ill be stored fo r decay, ce n tr ifuged to rem ove ex cess m oistu re , and placed in 55-g a l containers fo r shipment o ff-s ite . Concentrated waste (evaporator bottom s) w ill be m ixed with a m oistu re-absorben t m ateria l and placed in co n ta in e rs . A hydraulic b a ler w ill be used to com p act c o m p re ss ib le w astes fo r packaging and shipm ent o f f -s it e .

2 .3 .3 . L iquid w astes

Estimated flows and associated activities for the m ajor sources of liquid wastes are listed in Table XIX. These liquids, as indicated in the e a r lie r figu re , are segregated in fou r ca te g o r ie s : h igh - conductiv ity w a stes , low -con d u ctiv ity w aste , ch em ica l w aste and m isce lla n eou s w aste.

H igh-condu ctiv ity w astes a r is e fro m flo o r -d r a in su m ps, and are pumped into a d r a in -c o l le c to r . tank. This liquid is pumped through a p recoa t filte r to e ither o f two f lo o r -d ra in sam ple tanks. A fter sam pling and analysis the tank contents a re pumped into the c ircu la tin g -w a ter d isch arge tunnel.

180


T A B L E X V II. NINE M ILE POIN T STATIO N LIQUID W A STE DISPO SAL SYSTEM M AJOR COM PON EN TS

Component Quantity Capacity

1. Concentrated waste tank 1 5 000 gal

2 . C hem ical addition tank 1 600 gal

3. Drywell equipment sump 2 2 000 g a l3

4 . Drywell floor drain sump 1 2 000 gal

5. Electric boiler 1 8 000 lb /h

6. Filter aid tank 1 470 gal

7. Floor drain co lle cto r tank 1 10 000 gal

8. Floor drain filter 1 300 ga l/m in

9. Floor drain sam ple tank 2 20 000 gal a

10. Laundry drain tank 2 2 000 g a la

11. Precoat tank 1 560 gal

12. Reactor building equipment drain tank 1 5 000 gal

13. Reactor building floor drain sump 6 4 000 g a la

14. Turbine building equipment drain tank 1 1 000 gal

15. Turbine building floor drain sump 8 4 800 g a l3

16. W aste building equipment drain sump 1 2 000 gal

17. Waste building floor drain sump 3 3 000 g a la

18. Waste co lle cto r tank 1 25 000 gal

19. Waste co lle ctor filter 1 300 ga l/m in

20. Waste concentrator 1 12 ga l/m in

21. Waste dem ineralizer 1 300 ga l/m in

22. Waste neutralizer tank 1 15 000 gal

23. Waste sam ple tank 2 50 000 g a la

24. Waste surge tank 1 50 000 gal

a Total capacity

181


T A B L E X V III. E STIM A TE D ANNU AL SOLID W A STE A C C U M U LA TIO N AND A C T IV IT Y A T NINE M ILE POINT

Maximum shipment activity

Filter sludge

Normal volum e 1610 ft3 / yr 1 .1 C i/drum

Drums required 230 drums/yr

Spent resins

Normal volum e 755 ft3/y r 1 .1 C i/drum

Drums required 107 drums/yr

Concentrated waste

Normal volum e 0

Maximum volum e 400 ga l/d 2 .5 X 10"2 C i/drum

Drums required 9 drums/d

Dry wastes

Compressible wastes 500 ft3/y r a 0 .5 C i/y r

N on-com pressible wastes- 1000 ft3/y r 10 C i/y r

aBased on a 5 : 1 com paction ratio.

L ow -con d u ctiv ity w astes a r is in g fro m equ ipm ent- and pipe - drains are co llected in equipm ent-drain sumps o r tanks and pumped to the w a s te -co lle c to r tank. O ther, le s s frequent, sou rces include w aste effluents from the fu e l-p o o l coo lin g sy s te m , c lean u p - dem ineralizer system , containm ent-suppression cham ber, em ergency condensers, and backwash water from the condensate dem ineralizers. A waste surge tank is provided to co lle ct high w ater volum es which m ay be produced in frequently , and to p rov id e in terim stora g e fo r o ff-stan d ard batches w hich m ust be r e cy c le d through the w a s te - d isp osa l sy stem .

182


Liquid w astes from either the c o lle c to r tank o r surge tank are pumped through a h ig h -e ffic ien cy precoat filte r and a m ixed -bed dem ineralizer to either one of two waste sam ple tanks. A fter sam pling and ana lysis , the liquid is norm ally pumped to the condensate storage tank.

C hem ical w astes a r ise from the condensate dem in era lizer r e generation, laboratory sinks and equipment decontam ination drains, and a re co lle c te d in the w a s te -n eu tra lize r tank. If sam p les show a low activity , the waste can be neutralized and pumped through the f lo o r -d ra in p recoa t filte r into one o f the flo o r -d r a in sam ple tanks fo r d isp osa l to the circu la tin g -w a ter d isch arge tunnel. If sam ples o f the w aste show a high ra d ioa ctiv ity , the n eutra lized w aste is pumped to a waste concentrator, the bottoms from which are collected in the con cen tra ted -w aste tank. The concentrated w aste is m ixed in the w aste m ixer with an absorbent m ateria l and handled as solid w aste.

M isce llan eou s w astes co m p r ise th ose liqu ids fr o m laundry operations, personnel decontam ination or other rad ioactive liquids w hich m ight contain detergen ts . T hese are co lle c te d in laundry- waste tanks and, after sam pling and analysis, are norm ally pumped to the circu latin g-w ater d ischarge tunnel.

Sampling lines from each co lle ction tank, sam ple tank and up - and dow nstream from filte rs and d em in era lizer run to a sam ple station adjacent to the w aste d isp osa l fa cility con tro l room . L o ca l sam ple points are a lso provided . Data from sam ples a re record ed together with d isch arged w ater volu m e, and a continuous m on itor on the liquid w aste d ischarge line re cord s activity re leased and annunciates a high concentration alarm .

2 .3 .4 . Gaseous w astes

G aseous ra d ioa ctive w astes include a irb orn e p articu la tes as w ell as those gases vented from p rocess equipment. M ajor sou rces o f gaseou s w aste a ctiv ity in clude the m ain con d en ser a ir e je c t o r effluent and the turbine gland-sea l condenser exhaust. Building ventila tion exhausts con tribu te la rg e vo lu m es o f a ir to the w aste sy stem s but the le v e ls o f ra d ioa ctiv ity added a re in sign ifican t in com p a rison with the con d en ser a ir e je c to r a ctiv ity .

Estim ated flow s and associated activities fo r the m ajor sources o f gaseous w astes a re given in T ab le X X fo r the n o -fu e l fa ilu re condition. These activities re flect a 30-m in decay which takes place

183


T A B L E X IX . NINE M ILE P O IN T: E ST IM A TE D Q U A N TITIE S AND A C T IV IT IE S OF LIQUID R A D IO A C T IV E W ASTES

Liquid discharge (g a l/d ) A ctiv ity le ve l 0 iC i/m l)

Maximum Normal Maxim um Normal

1. L ow -conductivity liquid wastes

A . Dry w ell

(1) Recirculation pump seals 3600 2 1 0"1

(2) Valves (recirculation , seal, e t c .) - 1730 2 10"1

(3) M aintenance 4500 0 2 0

B. Reactor building

(1) Pump seal leakage (cleanup, rod drive, sludge) 990 2 X 1 0 '2 io - 3

(2) Valves (scram , feed , e t c .) - 1980 2 X 1 0 '1 ,10"2

(3) Resin transfer and filter drains - 1500 2 X 1 0 '5 lO"6

(4) Maintenance 600 O' 1 10‘ 2

C . Turbine building

(1) Feed pump seals 720 2 X 1 0 -5 10‘ 6

(2) Condensate pump seals - 720 2 X 1 0 "3 10‘ 4

(3) Sample drains - 720 2 X 1 0"4 1 0 "5

(4) M aintenance 2100 0 -

• O

.

t 1


TABLE X IX (cont.)

Liquid discharge (g a l/d ) A ctiv ity le ve l ((iC i/m l)

M axim um Normal Maxim um Normal

D. W aste disposal building

(1) O ff-gas drains 5 100

(2) Resin regeneration backwash 62 000 12 500 - 2 X 10-2

2 . H igh-conductivity liquid wastes

A . Drywell

(1) Floor drains , 500 2 X 10" 5 1 0 '6

(2) Control rod drive drains 2 500 2 X 1 0 '5 10‘ S

B. Reactor building

(1) Floor drains 2 000 10"2 1 0 '4

(2) M aintenance drains

'

(varies)

'

i o - 3


TABLE XIX (con t.)

Liquid discharge (g a l/d ) A ctiv ity leve l (n C i/m l)

Maximum Normal M axim um Normal

C . Turbine building

(1) Floor drains•

2 000 1 0 '5 1 0 '6

(2) Decontam ination 2 000 0 - 1

(3) Laboratory drains - 500 - 1 0 -4(4) Laundry 5 250 1 000 - < 1 0 '5

D. Waste disposal building

(1) Floor drains 2 000

1

1 0 '4

(2) Waste centrifuge - 3 000 - 1 0 -3

(3) Regeneration solutions - 8 700 4 X 1 0 '1 2 X 1 0 '3

(4) Decontam ination - 500 " 1

N ote; In som e cases m axim um quantities are not listed because the exact values are unknown. These m axim um values are expected to be close to the values listed as norm al.


T A B L E X X . NINE M ILE P O IN T : FLO W S AND A C T IV ITIE S OF M AJO R SOURCES OF GASEOUS A C T IV IT Y (NO F U E L FA IL U R E )

SourceDesign flow

(ft3/m in )Nuclides

Stack activity at 1779 MW (th)

(|iCi/s)

Condenser air 140 13 N 580

Ejectors 41A 7

Turbine gland seal 1900 13n 0 .6condensate exhaust 41a 0 .007

betw een the a ir e je c to r and the stack . F o r the fu e l fa ilu re con d ition , the design flow s a re the sam e as th ose lis ted in T ab le X X . T he ra d ioa ctiv ity em itted is co m p r ise d o f the sh o r t - l iv e d n o b le - g a s e s , predom inantly l38X e , 87K r , 88K r , 135X e and 136mX e . It is estim ated that an average re le a se rate o f 1 C i /s o f this m ixture fr o m the 2 4 0 -ft stack w ill not resu lt in exceed in g an ex p osu re at the n ea rest s ite boundary in e x c e s s o f 500 m i l l i r e m /y r . T he operating lim it fo r the stack re le a s e ra te w ill be estab lish ed by the operating lic e n ce issued by the A tom ic E nergy C om m ission .

A s indicated in the e a r lie r figu re , the a ir e je c to r exhaust draw s non -con den sib le gases fro m the m ain con denser through a 30-m in holdup pipe which perm its sh ort-liv ed activity to decay. T hose sh ortlived n o b le -g a s e s w hich d eca y into p articu la te daughters during this delay period are rem oved by two h ig h -e ffic ien cy filte rs p laced in s e r ie s n ear the end o f the holdup p ip e . An o f f -g a s m on itor is lo ca ted at the en tran ce to the holdup piping w hich au tom atica lly c lo s e s the va lve at the pipe exit i f the o f f -g a s a ctiv ity lim it is in excess of the sh ort-term re lea se lim it fo r 15-m in . A 15-m in delay is provided to perm it co rre ctiv e action to be taken in ord er to avoid u n n ecessa ry station shutdown. O ff-g a s p ip ing, v a lv es and f i lt e r housings a re designed to withstand the p r e s s u r e s gen erated by a p o s s ib le h y d rog en -oxy g en e x p lo s io n . The high p r e s s u r e o f su ch an explosion w ill a lso c lo se the valves upstream o f the holdup pipe.

187


The turbine g lan d-sea l condensate exhaust contains a relatively sm all amount o f steam leaking through the gland sea l and a r e la tively large amount of room a ir . Since activ ities are low these e x haust gases are delayed fo r 1.75 min to perm it the decay o f 16Nand lsO and are then exhausted to the stack.

The tu rb in e -re a c to r and w aste -bu ild in gs vent sy stem s are handled as indicated in F ig .4 . The en tire w a ste -b u ild in g exhaust is filtered b e fo re d isch a rge to the stack . A continuous m on itor in the stack indicates the activ ity re leased from the plant and alarm s in the con tro l room .

2 .3 ._5_._ New developm ents

In the new er boilin g -w ater rea ctors cu rrently under con stru ction o r in planning severa l changes are being introduced which have a significant bearing on waste managem ent. One is the substitution o f "P ow d ex"3 filter-d em in era lizers fo r the precoat filters and deep- bed d em in era lizers used in the waste treatm ent and rea ctor coolant cleanup system s and condensate d e m in e ra lize r . The other is the elim ination o f w aste con cen tra tors from the ch em ica l w a ste - treatm ent lin e . Each o f these is d iscu ssed further below .

The Powdex p rocess essentia lly rep laces the filter aid m aterial used in pressu re precoat filters with fine particle s ize ion exchange r e s in s . The Pow dex re s in is ex trem ely fin e , 90% being sm a lle r than 325 m esh . T h ere fore , the available surface area fo r exchange reaction is greatly in creased over that associated with conventional ion exchange re s in s . A dd ition a lly the fine p a rtic le s iz e p rov id es an ex ce llen t filtra tion cap ab ility . B eca u se o f the high u sab le e x change cap acity p er unit w eight, the r e s in is not regen era ted but d isposed of after exhaustion. This elim inates chem ical w astes from the regeneration p ro ce ss but does resu lt in som e in crea se in resin d isch a rge , p articu larly from the condensate d em in era lizers which in conventional fo rm m ay have a s e rv ice life o f one o r two y e a rs . T h ese f i lte r -d e m in e r a liz e r units have been em p loyed in con v en tiona l steam -p lan t condensate d e m in e ra lize rs but have not as yet dem onstrated their perform ance in nuclear station operation. Howe v er , they do o ffe r con sid erab le p rom ise fo r elim ination of the separate fi lte r and d em in era lizer units.

s Trade-mark o f Graver Water Conditioning Com pany.

188


As indicated by the operating experience fo r the BWRs described in the preced in g se ction s , there has been essen tia lly no re q u ire ment fo r the u se o f w aste evap ora tors (o r con cen tra tors ) fo r the decontam ination o f ch em ica l w astes. This highly favourable waste operating exp erien ce has led m ost o f the new er units to elim inate these w aste concentrators from their design . It w ill be sev era l y ea rs b e fo re the e ffe c tiv en ess o f th ese new sy stem s can be determ in ed .

3. P R E SSU R IZE D -W A TE R REACTORS

3 .1 . Introduction and gen era l d iscu ssion

In a p ressu rized -w a ter re a c to r (PW R) the ligh t-w ater coolant is pumped under high p ressu re through the rea cto r c o re w here heat is tra n s fe rred fro m the fu e l r o d s . The re a c to r coolant p a sses to a steam gen erator w here heat is tra n sferred to a secon d ary w ater system b e fo re the return o f the rea ctor coolant to the bottom of the reactor co re . The secondary water is boiled in the steam generator, and the steam passes to the turbine w here it condenses and is r e turned to the steam generator. The reactor coolant usually contains 20 to 40 cm 3 o f hydrogen per k ilogram o f w ater to inhibit rad iolytic d ecom position o f the re a cto r coolant. C ontrol of pH m ay be e x e rc ise d using enriched lithium-7 hydroxide. In addition, all the pressu rized - water rea ctor plants operating at the present tim e em ploy b or ic acid in the p r im a ry coolant system as a soluble neutron poison both to con tro l lon g -te rm changes in reactiv ity o f the c o r e during plant operation , as w e ll as fo r co ld shutdown reactiv ity con tro l.

R eactor coolant quality is usually maintained by passing a por - tion o f the m ain coolant flow through a bypass purification system . T his in clu d es a tank in w hich hydrogen and oth er ch e m ica ls m ay be added and other gases are rem oved ; and a se r ie s o f d em in era liz e r s and f i lte r s to re m o v e suspended and d is s o lv e d im p u r it ie s . L a rg e changes in b oron con cen tra tion in the r e a c to r coo lan t (fo r exam ple, follow ing a cold shutdown) are usually achieved by dilution with pure makeup w ater ra th er than by rem ov a l o f boron by anion exch an gers.

Secondary w ater quality is m aintained e ither by fu ll-flo w d e - m in era liza tion o f the condensate o r by blow dow n o f the secon d a ry s id e o f the steam g e n e ra to r . S ince the stea m g e n e ra to r is o la te s

1 8 9


the r e a c to r coolant fro m the turbine c y c le the se con d a ry w ater is n orm ally not contam inated with ra d ioa ctiv ity .

Under n orm al operation c o r ro s io n products and m inor c o n cen trations o f f is s io n g a ses fr o m "tra m p " uranium constitute the m ain so u rce o f ra d ioa ctiv e w astes fro m P W R s. T ritiu m is a lso produced from the boron and lithium in the coolant. The co rro s io n produ cts w ill v a ry depending upon the m a ter ia ls u sed in the c o n struction o f the rea ctor coolant loop , but w ill usually contain ra d io active iso top es o f coba lt, m anganese, iron and ch rom iu m . Liquid w astes a r ise from re a c to r coolant expansion on startup, from the d ilution o f r e a c to r coolant w ater fo r b oron con cen tration co n tro l, fr o m sam plin g a ct iv it ie s and fr o m equipm ent and p r o c e s s le a k s , laboratory operations, decontam ination activ ities , etc . Since large volum es o f gases are not involved, a ll p ressu rized -w a ter rea ctors provide gas storage tanks to perm it the retention fo r decay o f those radioactive gases extracted from the rea cto r -coo la n t system or the w a s te -p ro ce ss system . The em iss ion s of w aste gases from PW Rs gen era lly a re n u m erica lly sm a lle r in quantity than those re lea sed fro m BW Rs although they con s is t la rg e ly o f the lo n g -liv e d fis s io n gases rather than the predom inantly sh ort-liv ed fiss ion gases p r o duced by B W R s. Since the gases can be s tored th e ir re le a se can , if needed, be tim ed to co in cide with the m e te o ro lo g ica l conditions m ost favourable fo r d isp ers ion .

3 .2 . S p ecific PWR exp erien ce

T he data below on the Indian P oint and Y ankee r e a c to r s a re again la rg e ly taken fro m the w ork o f B lom eke and H arrington [1] with additional data that have b ecom e availab le s in ce th e ir rep ort w as issu ed .

3 .2 .1 . Indian Point

3 .2 .1 .1 . G en era l

C onsolidated E d is o n 's Indian Point N o .l . n u c le a r -s te a m - generating plant is located on the east bank of the Hudson River about 24 m iles north o f New Y ork City [17, 18]. The rea ctor is a B abcock and W ilcox Com pany p re ssu r ize d -w a te r type that was designed to use a m ixture o f UO2 and ThC>2 pellets as fuel, plus an o il-f ire d su perheater, to produce 275 MW o f e le c t r ic ity . The re a c to r f ir s t

190


achieved critica lity on 2 August 1962, and was placed in com m ercia l operation on 1 O ctober 1962. This unit has produced 5 784 000 MWh o f e le c t r ic ity up to 31 D ecem b er 1967, w ith a ca p a city fa c to r o f about 0.55 [ 10].

The plant ch a ra c te r is t ic s that a ffect the nature o f the r a d io a ctive w astes a re su m m a rized in T ab le X X I. T he f ir s t c o r e fuel w as a m ixtu re o f 1100 kg o f uranium as 93%-en rich ed 1235 UO2 and17 000 kg o f thorium as Th02 p ellets . Type 304 stain less stee l was used as the cladding m ateria l. This was rep laced in late 1965 with a slightly enriched UO2 core designed and supplied by Westinghouse; a third o f this fuel was rep laced in late 1966. F igure 5 is a s im p lified flow sh eet show ing the s o u r c e s , treatm ent and d isp os ition o f rad ioactive w aste at Indian Point. The activ ities and volum es are design estim ates. The waste p rocesses and equipment are sum m arized in Table XXII.

T A B L E X X I. P L A N T C H A R A C TE R ISTIC S T H A T A F F E C T R A D IO A C T IV E W A ST E G E N E RA TIO N A T INDIAN POIN T

Coolant Light water

C oolant c y c le Pressurized water

T ype o f fuel T y p e -304 stainless-steel-clad ,93'7o- enriched UO2 and TI1O 2 pellets

Fuel failures Nil

Control rod m aterials Stainless-steel clad A g-In -C d

Other m aterials exposed to primary coolant

Austenitic stainless-steel primary equipment

Z irca loy -2 (fuel structure)

Stellite (va lve seats)

Coolant purification methods Ion exchange, filtration, degasification

Additives to primary system H2 , lithium hydroxide, boric a cid , ammonium hydroxide

191


T A B L E X X II. W A STE PRO CESSES AND EQ U IPM EN T USED A T INDIAN POIN T

Gaseous wastes:

Holdup tim e Discharge rate Dilution available Stack height FilterC ata lytic hydrogen recom biner

Liquid wastes:

FiltersDem ineralizersConcentrator

Tank volum e Discharge rate Dilution available

Solid wastes:

Drumming station Interim storage

120 d3 ftV m in (STP)280 000 ft3/m in (STP)400 ftAbsolute

NoneOne 1 2 -g a l/m in m ixed -bed unit One 1 2 -g a l/m in natural-circulation

type 350 000 gal 25 ga l/m in 280 000 ga l/m in

Includes two concrete mixers Four 1500-gal resin storage tanks

and two 335 0 -gal evaporator concentrate tanks

The Indian Point rea ctor operates under AEC L icen se No. D P R -5 , and ra d io a ct iv e -w a ste d is ch a rg e s a re lim ited by regu la tion s set forth in the lice n ce and in T itle 10, Code o f F e d e ra l R egu lations, P art 20. In the c a s e o f g a seou s d is c h a r g e s , p a rticu la tes and halogens with h a lf-lives longer than eight days shall not be released at an instantaneous rate (C i/s ) exceed in g 2.4 X 104 m ultip lied by M PC; and the rate o f re le a se (C i/s ) averaged ov er one y ea r shall not exceed 2.4 X 103 m ultiplied by M PC. A ll other radioactive i s o topes in gaseous waste shall not be released at an instantaneous rate (C i /s ) exceed in g 1.7 X 107 tim es M P C , and at an annual average rate (C i/s ) not to exceed 1.7 X 10® tim es M PC. MPC is defined as the m axim um p erm iss ib le con cen tration , in n C i /c m 3, as listed in A ppendix B , T able II, Colum n 1 of T itle 10, C ode o f F ed era l

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Regulations, Part 20. If 131I is assum ed to be controlling , an MPC o f 1 X 1 0 '10is used fo r halogens and p articu la tes , and the in stan taneous and annual average a llow able r e le a s e ra tes a re 2.4 and 0 .2 4 /u C i/s re sp e c t iv e ly . A pplying an M PC o f 1 X 10- I ° n C i /c m 3, the instantaneous and annual average a llow able re le a s e ra tes a re 500 000 and 50 000 nCi / s resp ective ly . No radioactive liquid wastes having concentrations in excess o f those specified in T itle 10, Code o f F ederal Regulations, Part 20, may be d ischarged from the plant.

3.2.1 .2 . Solid wastes

3 .2 .1 .2 .1 . Sources

The so lid -w aste system essentia lly con sists o f six surge tanks and a con crete packaging station. F our o f the tanks are used to r e ce ive depleted ion-exchange res in s , and two are used fo r evaporator con cen trates. Ion-exchange res in s a r ise from the rea cto r -coo la n t purification system , the spent-fuel storage-poo l purification system and the liqu id-w aste system .

M ateria ls sh ipped, in addition to re s in s and evap ora tor c o n centrates, include contaminated too ls , ra g s , p la stics , instruments and equipment.

3 .2 .1 .2 .2 . E xperience

F rom August 1962 to the end of September 1967, 648 steel drums o f m isce llaneou s radioactive rubbish w ere shipped o ff site fo r d is posa l. The drum s represented a total volum e o f about 4800 ft3 and contained about 3.5 C i o f a ctiv ity . S ince u se of the w aste c o n centrator has not been requ ired (see below ), no evaporator co n centrates have been produced. How ever, the waste volum e shipped has exceeded the quantities orig ina lly estim ated fo r m isce llaneou s w astes by a lm ost a fa cto r o f 10.

3 .2 .1 .3 . Liquid w astes

3 .2 .1 .3 .1 . Sources

The liquid-w aste system is designed to rece iv e , store and p ro c e ss a ll plant ra d ioa ctive -liq u id w astes so that they m ay e ither be returned to the p rim a ry coolant system or d isch arged through the con denser d isch a rge tunnel to the Hudson R iver.

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T A B L E X X III. R A D IO A C T IV E GASEOUS AND LIQUID W ASTE S DISCH ARGED A T INDIAN P O IN T 2

Period

Gaseous releases Liquid discharges Gross electr ic ity

generated (MWh)Average rate

(pC i/s)Volum e (1 0 3 gal)

Concentration(yC i/cm 3)c

T otal(C i)

2 A u g .1962 - 31 Jan .1963 0 2 337 1 .5 1 x 10’ 5 0 ,1 3 4 185 300

1 Feb. 1963 - 31 Jul. 1963 0 2 138 3 .2 8 X 1 0 "e 0 ,0 2 7 ^ 543 400

1 A u g .1963 - 31 Jan .1964 0 .1 3 3 630 9 .8 2 x 10 '* 0 .135 444 700

1 F e b .1964 - 31 Jul. 1964 0.001 1 478 4. 50 X 30‘ 4 2. 52 14 300

1 A u g .1964 - 31 Ja n .1965 0 .7 1 11 222 2 .4 9 x 10’ 4 10 .56 623 100

1 F e b .1965 - 31 J u l.1965 0 .5 8 4 176 8 .3 9 x 10‘ 4 13 .26 e 634 000

1 A u g .1965 - 31 Ja n .1966 1.52 1 696 2 .0 4 x 10*s 13 .0 9 f 364 400

1 F e b .1966 - 30 Sep. 1966 1.62 10 352 1. 80 X 10"4 7 .0 0 8 679 900

1 O ct. 1966 - 31 Mar. 1967 0 .3 6 6 466 4. 62 X 1 0 *s 5 .9 9 h 1 102 000

1 Apr. 1967 - 30 Sep. 1967 1.10 7 569 6 .3 9 X 1 0 '8 4 .7 9 1 1 120 000

3 C om piled from Consolidated Edison C o . Sem i-Annual Operations Reports for Indian Point Station [1 9 ] .k Volum e before dilution with condenser-circulating water. c Average concentration, before dilution.^ In addition, 34 fiCi were released in sewage to treatment plant.e In addition, 233 C i o f tritium were released at an average concentration, after dilution , o f 3 x 10“ 6 p C i/c m 3 . In addition , 258 C i o f tritium were released at an average concentration , after d ilu tion , o f 4 x 10'® j iC i /c m 3.

8 In addition, approxim ately 118 C i o f tritium were released at an average concentration , after dilution , o f 3 x 10*T p C i /c m s.^ In addition , approxim ately 107 C i o f tritium were released at an average concentration , after dilution , o f 4 x 10’ T p C i/c m s .* In addition, approxim ately 178 C i o f tritium were released at an average concentration , after dilution , o f 6 .5 x 10*7 p C i /c m 5.


Although the waste types are categorized in four groups (reactor plant liqu ids, ch em ica l w a stes , sp ec ia l w astes and n orm a lly non rad ioactive w astes), a ll except the laundry w astes a re co lle cted in four 75 000-gal collection tanks. These tanks provide tim e for decay o f sh ort-liv ed iso top es . The w astes from these tanks m ay be g a s - stripped , evaporated, dem inera lized and co lle cted in two 3350-ga l m on itorin g tanks. The ev a p ora tor bottom s a re tra n s fe r re d to a sludge storage tank, from which they may be m ixed with cem ent and packaged in dru m s. It is p o ss ib le to bypass the gas s tr ip p er , evaporator and d em in era lize rs , and to d isch arge w aste d ire ct ly to the con denser d isch arge tunnel.

3 .2 .1 .3 .2 . E xperience

T able XXIII su m m a rizes the w astes d isch a rged fro m Indian P oin t-I [19]. In actual operation it has been u n n ecessary to evaporate liquid waste to a concentrate fo r concreting and encapsulation. The evaporator has been operated, but only to test its operational characte r is t ic s . E ssen tia lly a ll liqu ids reach in g the w aste system have been d isch a rged to the r iv e r on the b a sis o f unidentified is o to p e s , although p e r io d ic iso to p ic an a lyses have in d icated that the lim it , based on actual iso top es p resen t, would be 2 X 10-6 ^ C i /c m 3 [1 ].

3 . 2 . 1 . 4 . G aseous wastes

3 . 2 . 1 . 4 . 1 . Sources

The waste gas system s include equipment to rece iv e , p ro cess , store fo r decay and safely discharge radioactive gas m ixtures. The p rim e sou rce o f w aste is d isso lv ed hydrogen and tra ce s o f r a d io active m ateria l, w hich are re le a se d when the p r im a ry coolan t is cooled and depressurized . The prim ary loop de -aera tor is the point o f m ajor re lease ; a second point o f release is the waste system gas str ip per and the evaporator. The gases are passed through a con den ser and a surge tank to a ca ta lytic com bin er. In the com biner, the water vapour from the H2- 0 2 com bination is condensed and separated, and the m o is tu re -free gas enters the fiss ion -g a s surge tank. T w o p a ra lle l c o m p r e s s o r s pump the gases from this tank to fou r surge tanks designed to p rov ide storage fo r 120 d. R elease to the atm osphere via the stack is at a con tro lled rate through a 5 5 -cm 3/m in absolute filte r .

195


P rim a ry coolan t system equipm ent is vented to a sw eep -g a s system , which is also used to con tro l the concentration o f hydrogen in the w aste tanks and piping. This system contains two 5 0 0 0 -ft3 surge tanks that can be pressurized to 75 lb /in 2 gauge. The gas can be rec ircu la ted from these surge tanks through the cata lytic c o m biner as needed, o r it can be released through the ca rb on -stee l gas holdup tanks.

3. 2. 1.4. 2. E xperience

The w aste gas r e le a s e s a re su m m arized in T ab le XXIII, In operation to date it has been unnecessary to holdup the waste gases fo r decay. A ll gases reaching the fiss ion -g a s surge tank have been released through the stack without com pression and storage. Thus, the com p ressors have been run only fo r testing purposes.

3 . 2 . 2 . Yankee

3. 2. 2. 1. General

The Yankee A tom ic E lectr ica l Com pany's nuclear rea ctor [20, 21] is located in Rowe, M assachusetts, on the east bank o f the D eerfield R iver, 10 m iles east o f North Adam s. The W estinghouse p ressu rized -w a ter rea ctor has had its capacity in creased from 120 to 185 MW (e) in the course o f m ore than five years of operation. Up to 31 D ecem b er 1967, 7 711 000 MWh o f e le c t r ic ity had beengenerated.

The plant ch aracteristics that affect the nature o f the radioactive wastes are sum m arized in Table XXIV. The first core consisted o f 3 .4% -enriched UO2 pellets clad with 0. 021 -in -th ick type 348 stainless steel. Later cores are sim ilar except for the use of fuel having a slightly higher enrichm ent. The m ateria ls subject to c o r ro s io n by the ligh t-w ater p rim ary coolant con s is t p r im a rily o f types 304, 348 and 410 stainless steel, but also include Z ir c a lo y -2 con trol rod follow ers, nickel-phosphorus braze m aterial, Inconel, Inconel X and n ickel. The prim ary coolant purification system , consisting o f cartridge filte rs and m ixed-bed d em inera lizers , can p rocess up to 100 g a l/m in o f the c ircu la tin g w ater. M eans are prov ided to add ch em ica ls , such as hydrazine, NH3, and elem ental hydrogen, to reduce d isso lved oxygen in the p rim ary coolant and to con tro l pH. B oric acid is added to the prim ary coolant to achieve chem ical shut-

1 9 6


TABLE X X IV . PLANT CHARACTERISTICS THAT A F F E C T RADIOACTIVE W ASTE GENERATION AT YANKEE

Coolant

Coolant cycle

Type of fuel

Fuel failures

Control rod material

Other material exposed to primary coolant


Light water

Pressurized water

Type-348 stainless-steel-clad UOz

1%, design; < 0 .V ’jo, actual

Ni-plated or stainless-steel-clad Ag-In-Cd alloy

Type-304 stainless-steel reactor vessel and piping; Zircaloy-2 in control rod followers; type-410 stainless steel, Inconel in miscellaneous reactor components; Ni-P braze from fuel bundle

Up to 100 gal/min demineralization (non-regenerative), and filtration

Additives to primary aqueous system

During operation During cold shutdown

Other chemicals

NH j, Hj, hydrazine Soluble boron

Boron shim employed

down and for chemical shim control. Initially, Yankee operated with boron used only for cold shutdown, and it was subsequently removed at reactor startup by dilution followed by anion exchange. Chemical shim as a supplement to mechanical control rod operation requires a gradual reduction of the boron during operation to compensate for burnup.

Figure 6 is a simplified flowsheet showing the sources, treatment, and disposition of radioactive waste at Yankee. This figure is a revision from an earlier study by the Edison Electric Institute 12) . The waste plant processes and equipment are summarized in Table XXV.

1 9 7


TABLE X X V . CAPACITIES AND RATINGS, OF WASTE MANAGEMENT EQUIPMENT AT YANKEE

Tanks Net operating capacity

1 Primary building sump tank (gal) 484

1 Gravity drain tank (gal) 4 700

2 Monitored waste tanks (gal each) 1 370

1 Primary drain collecting tank (gal) 7 500

1 Waste holdup tank (gal) 75 000

1 Activity dilution decay tank (gal) 75 000

1 Distillate accumulator (gal) 60

2 Test tanks (gal each) 8 040

1 Compressor "knockout” drum

1 Waste gas surge tank (ft3) 4 160

3 Gas decay drums (ft3 each) 60

1 Primary water storage tank (gal) 135 000

1 Dilute caustic storage tank (gal) 235

1 Ash dewatering sump (gal) 100

1 Cyclone separator (gal) 110

Other equipment Rating (each)

16 Pumps (gal^nin each) 10 to 200

2 Waste gas compressors (ft3/min (STP)) 20

1 Disposable filter unit (gal/min) 6

1 Incinerator, lb/h 40

3 Roto-Clone, lb/h 480

1 Evaporator (gal/min) 6

1 9 8


The principal changes that have been made in the original system include the replacement of the electrode evaporator by a steam - heated, forced-circulation reboiler, the elimination of the gas stripper and the installation of additional cartridge filters.

Radioactive discharges are limited by regulations set forth in Title 10, Code of Federal Regulations, Part 20. The plant is licensed, No.DPR-3, to discharge gaseous wastes in concentrations not in excess of 1000 times the limits specified in Appendix B, Table II, Title 10, Code of Federal Regulations, Part 20. Modified MPCs selected for use under controlled conditions based on periodic isotopic analysis of the stream under consideration have established:

(1) For "unrestricted area" waterborne activity, with radioiodine controlling, the MPC is 8 X 1CT6 pC i/cm 3.

(2) For "unrestricted area" airborne activity, with noble-gas radioisotopes controlling, the MPC for air ejector effluent is 3 X 10"7 |UCi/cm3 .

No cost data are available for this plant.


3 .2 .2 .2 .1 . Sources

Combustible solids, such as absorbent paper, cloth and shoe coverings, are transferred to the waste-disposal building in com bustible fibre drums. The drums are re-used until contaminated. These solids are burned in an incinerator using free vortex flow of combustion air over the grate. An 80-lb charge is burned in about2 h. The off-gases are processed as described below. The residue, after complete burning, is dropped through the cone bottom into an open-top 55-gal steel drum half-filled with water. When the drum contains about 18 batches o f ash, two sprays inside the head are operated to thoroughly wet the ash. Liquid is decanted from the settled ash so that the volume remaining in the drum will solidify when mixed with cement. The decanted liquid is pumped to the gravity drain tank. The drums of solidified ash are shipped by a l i censed contractor to appropriate burial sites.

Filter cartridges and contaminated equipment items are, if possible, immobilized in concrete in 55-gal drums for shipment to the burial site. Large items that cannot be decontaminated are shielded for temporary storage until special shipment can be arranged.

♦

1 9 9


200 T A B L E X X V I. R A D IO A C T IV IT Y R E L E A SE D IN YA N K E E W A ST E E F F L U E N T S3Gaseous Liquid Solid

PeriodTotal

activity(C i)b

Percent o f lim it0

Volume*1(1 0 3gal)

Activ ity(m C i)b

Percent o f MPC®

Number o f containers*

T otalactiv ity

(C i)

Gross electr ic ity generated (MWh)

1961 0 .0 0 2 . Nil 592 8 .0 0 .0 1 70 0 .1

1962 2 1 .7 0 .0 3 867 1 .6 0 .006 517 1 .5

1963 7 .4 0 .0 1 684 3 .5 0 .0 0 4 656 60 0001

1964 1 .0 0 .001 } 2.0* 507 60 000m 1257 000

1965 1.38 0.002 829 29.3J 0 .0 3 669 319 1028 000

1966 2.4h 0 .0 0 4 1284 36. l k 0 .0 2 451 492 1371000

1967 2.3 n 0.004 2568 5 5 .8 ° 827 4 .8P 1348 424

* Com piled from Refs [2 2 ,2 3 ] .Excluding tritium.

^Based on a m axim um permissible concentration at discharge o f 3 x 10“ 4 f iC i/cm 3 , averaged over 12 months.Volum e before 4 7 0 0 -to - l dilution with condenser water.Average percent o f MPC at tim es o f discharge, based on permissible concentrations o f identified isotopes as given in Colum n 2 , Table 2 , Appendix B o f 10 CFR 20 after a 4 7 0 0 -to - l dilution with condenser water (140 000 g a l/m in ).Containers o f solid waste average about 7 .5 ft9 each , in volum e.

®In addition, 16 C i o f tritium were released.. In addition, 11 C i o f tritium were released.. In addition, 8 m Ci o f unidentified activity leaked to the soil from the ion exchange pit.^ln addition, 1290 C i o f tritium were released, and 31 m C i o f unidentified isotopes and 13 C i o f SH leaked to the soil from the ion exchange pit.j l n addition, 1922 C i o f tritium were released.

A ll but 14 C i associated with 24 core I control rods shipped in one container.A ll but -1 0 0 C i associated with 24 core 11 control rods shipped in one container.In addition, 15 C i o f tritium and 139 jiCi o f beta particulates were released.In addition, 1463 C i o f tritium were released.

^In addition, 4 casks containing a total o f 37 000 C i comprising control rod assemblies and test specim ens were shipped in 1967.


The spent resins from the two mixed-bed primary-coolant purification dem ineralizers, the two boron-rem oval anion exchangers andthefuelpit purification mixed-be'd exchanger are not regenerated. Originally, when the entire ion exchanger units were depleted, they were submerged in a w ater-filled pit for decay until o ff-s ite shipment. More recently, the resins have been sluiced directly to suitable containers for burial.

3 .2 .2 .2 .2 . Experience

The shipments of solid wastes are listed in Table XXVI and are the largest for any reactor considered in this study, mainly because of the evaporation treatment given to the liquid wastes. Most of the solid wastes consist o f concreted evaporator bottom s; only about two drums of incinerator ashes are produced per year. Two shipments of control rods in 1963 and 1964 account for most of the activity.

3. 2. 2. 3. Liquid wastes

3 .2 .2 .3 .1 . Sources

Liquid wastes are segregated into hydrogen- and air-bearing wastes. The hydrogen-bearing reactor plant effluents resulting from activity and boron dilution o f the prim ary coolant circuit constitute the m ajor volume and source of activity in the liquid-waste system. A ir-bearing wastes include decontamination fluids, the incinerator system 's Rotoclone drain, laboratory sinks and secondary coolant system blowdown.

Provisions are available to pass all hydrogen-bearing liquids through a filter and all boron-free hydrogen-bearing liquids through a m ixed-bed dem ineralizer before introduction to the liquid-waste system . This alternative is not shown in F ig. 6. The filter may also be used to rem ove particulates from the evaporator bottoms.

The activity-dilution waste is pumped from a low -pressure surge tank (not shown) to the appropriate decay tank, boron dilution waste is pumped from the low-pressure surge tank to the waste holdup tank, and the miscellaneous wastes that flow by gravity to the primary drain collecting tank are pumped to the waste holdup tank. The transfers are batch operations. If necessary, the activity-dilution liquid may be stored for as much as 30 d for decay before further

201


treatment. Depending on their activity content, these hydrogen- bearing liquids are either discharged through a disposable-cartridge filter to the condenser cooling-w ater effluent or processed by evaporation methods.

The air-bearing wastes are collected in two 1370-gal monitored tanks. Depending on activity content, these wastes may be d ischarged at a controlled rate to the main condenser cooling-w ater discharge line o r transferred to the gravity drain tank.

Liquids in the gravity drain, activity-dilution, and waste holdup tanks are processed through the evaporator in separate batches to prevent possible air-hydrogen explosive m ixtures.

These batches of waste are passed through a cartridge filter to a forced-circulation evaporator. The condensate is pumped to either o f two test tanks. A subsequent analysis serves as the basis for transfer to prim ary-w ater makeup or discharge to the river after dilution with the 140 000-gal/m in condenser cooling-water flow. All transfers to the river are at a rate of 30 gal/m in. The evaporator concentration factor is controlled by monitoring the contents of the evaporator. The concentrated liquid, when mixed with concrete and solidified in a 55-gal steel drum, may be shipped off site within 60d.

3. 2. 2. 3. 2. Experience

Liquid waste discharges are summarized in Table XXVI. The radioactivity released, not including tritium, in liquid wastes to the D eerfield R iver averages 14 m C i/y r, and it was discharged at concentrations in the range 0. 004 to 0. 03% of MPC (taken as 8 X 10'6 pC i/cm 3). The wastes were discharged at 0. 13% of the limit for continuous release o f unidentified isotopes, averaged over 12 consecutive months.

In 1965, 1966 and 1967 substantial quantities of tritium areknown to have been released in the liquid wastes at average concentrations of 5 - 7X10~6 |uCi/cm3. The total tritium released is about0.2% of the annual allowable lim it stipulated in T itle 10, Code of Federal Regulations, Part 20.

3. 2. 2. 4. Gaseous wastes

3. 2 .2 .4 .1 . Sources

The bulk of activity processed in and discharged by the off-gas system is released from the liquids discharged to the waste-disposal

202


plant from the primary coolant system. Most of the gaseous radioisotopes and hydrogen are released when the coolant is depressurized; the remaining gases are released in the evaporator. A com pletely closed header system, serving all reactor-liquid-effluent receivers, co llects all hydrogen-bearing gases. This system is purged with nitrogen before waste is discharged from the primary plant to prevent the accumulation of explosive hydrogen-air m ixtures. The gases are com pressed and collected in a 41S0-ft3 steel surge drum. Three 60-ft3 decay drums may be filled from the surge drum for additional decay before release to the prim ary 150-ft stack. The gases are released at a controlled rate through deep-bed filters, mixed with 15 000 ft3 /m in of air, and discharged through the stack. The con tainment purge system exhausts directly to the primary stack without filtration and is operated only when personnel access is required. The internal air filtration system controls the normal content of a irborne particulate activity within the containment building. The total flow of purge and dilutions air is 15 000 ft3/min. The main condenser a ir-e jection effluent is also exhausted to the prim ary building exhaust fan.

The flue gases from the incinerator are passed through a wet- gas scrubber, filtered through a deep-bed glass-w ool filter, and discharged through a separate stack.


Gaseous release history is summarized in Table XXVI. The activity released from Yankee is generally less than that of any other plant studied. The total activity (exclusive o f tritium) discharged in gaseous wastes has been typically less than 10 C i/y r , and has never been greater than 0. 03% of the limit. In the case of Yankee, a dilution factor of 1000 for atmospheric diffusion, averaged over the entire year, is used with an MPC of 3 X 10"1 /uCi/cm3 in establishing the limit.

3 • ? 1 3 • _ San Onofre

3 .2 .3 . 1. General [24]

The San Onofre Nuclear Generating Station, owned by the Southern California Edison Company and the San Diego Gas and Electr ic Company, is located on an 84 -acre site on the P acific coast

2 0 3


about 2\ m iles from the nearest edge of San Clemente, California. Unit 1 at this site is the first of the new generation of power reactors, representing an increase of more than a factor of two in power level over pressurized-water reactors previously operating. The reactor is a Westinghouse pressurized-water type using stainless-steel clad enriched uranium dioxide pellet fuel assem blies with a rating o f 450 000 gross e lectr ic kW. The reactor first achieved critica lity on 14 June 1967, and entered com m ercial operation 1 January 1968. This unit has produced 386 500 MWh of electricity up to 31 December 1967, with an availability of 62. 5% [10] .

Plant characteristics that affect the nature of the radioactive wastes are sum m arized in Table XXVII. R eactor coolant quality is controlled by operation of the chem ical and volume control system, which includes m ixed-bed purification dem ineralizers designed to operate at !90-gal/m in. Each dem ineralizer contains 25 ft3 of nuclear grade resin. Hydrogen and potassium hydroxide are added for corros ion and pH control. An additional purification system cation exchanger, containing 11 ft3 of hydrogen-form resin, is p rovided for removal of excess 7Li produced by the 10B(n, a-)7Li reaction.

TABLE XXVII. PLANT CHARACTERISTICS THAT A FF E C T RADIOACTIVE WASTE GENERATION AT SAN ONOFRE

Coolant

Coolant cycle

Type of fuel

Fuel failures

Control rod material

Other material exposed to reactor coolant


Additives to reactor coolant

Light water

Pressurized water — chemical shim

Type-304 stainless-steel-clad UOa

1% design

Stainless-steel-clad Ag-In-Cd alloy

Austenitic stainless steels, Inconel

Filtration (ceramic filter medium),demineralization (25 ft3 each, non- regenerative)

Hydrogen, boric acid, lithium hydroxide, hydrazine (after cold shutdown)

2 0 4


Figure 7 is a sim plified flowsheet showing the sources, treatment and disposition of radioactive waste at San Onofre. The design of the radioactive-waste disposal system is adequate to permit plant operation with up to 1% of the fuel rods having minor cladding defects. For nominal cladding defects of 1% the maximum coolant activity at 140°F was calculated to be approxim ately 280 juCi/m l distributed approxim ately as fo llow s : (a) corros ion products, 0 .05 ju C i/m l; (b) gaseous fission products (except tritium), 232 pCi/m l; (c) non- gaseous fission products (except tritium), 35 juCi/ml; (d) tritium, 13. 5 juCi/ml.

The San Onofre reactor operates under License D P R -13 from the US Atomic Energy Com m ission which includes restrictions on radioactive waste releases. F or liquid wastes, these require that the radioactivity concentration in the circulating-w ater system be maintained below the lim its specified in Title 10, Code o f Federal Regulations, Part 2 0, with one of two pumps operating. The circu lating-water pumps are rated at 175 000-gal/m in each, and an interlock exists which automatically halts liquid release to the circulating- water system when neither pump is operating.

Gaseous wastes are controlled to maintain o ff-s ite radiation exposures below those specified in Title 10, Code of Federal Regulations, Part 20. Based on the average m eteorology at the site, a 40 000 ft3/™ !!1 air discharge and a calculated dilution of 5.5 X 1 O'6 s /m 3, the following stack gas concentrations are established as operating lim its :

131

STACK CONCENTRATION, nCi/cm ^

Running Maximum averaged over365-day average any one hour

I 9.7 X 1 0 '1 9 .7 X 1 0 "6

Otherunidentified 9 .7 X 1 0 '7 9 .7 X 1 0 -6isotopes

Noble gases 2.86 X 1 0 '2 9 .0 X IO "2

2 0 5


An interlock is provided which automatically halts gaseous re lease from the radwaste system if neither fan is operating.


3. 2. 3. 2. 1. Sources

All com pressible solid wastes such as contaminated clothing, rags, wiping towels, paper, gloves and shoe coverings, are com pressed into drums by a hydraulic baler. Non-com pressible solids such as wood, metal, glass, p lastics, concrete and ceram ics are put into drums.

These drums and other contaminated item s too large for packaging will be stored in a concrete storage vault in the reactor auxiliary building for radioactive decay and future disposal off site. The storage vault has a 2000 ft3 capacity to provide an estim ated minimum of three years storage of all solid wastes except spent resins.

Spent resins from the chemical and,volume control system and from the liquid-radioactive-waste-treatm ent system are sluiced to a spent-resin storage tank of 700 ft3 capacity for radioactive decay and future disposal o ff-s ite .


No report of solid-waste disposal has been made up to 31 December1967.

3. 2. 3. 3. Liquid wastes

3 .2 .3 .3 . 1. Sources

The majority of radioactive liquid wastes derive from dilution of the boron-containing reactor coolant during normal plant operation. In addition, further quantities are generated by hot or cold shutdowns o f the reactor during which the boron concentration in the coolant must be adjusted, with a consequent large quantity of liquid waste. Startup from a cold condition results in therm al expansion o f the coolant, and the excess also becom es liquid waste.

Because shutdowns generate large volumes of radioactive liquid wastes, the design o f the waste disposal system was established on

2 0 6


the basis of the following postulated reactor operating cycle: (a) one refuelling startup for ea.ch core cy cle ; (b) one cold shutdown and restart immediately following initial full power operation of the core cycle; (c) four hot shutdowns and restarts evenly distributed throughout each core cycle with the provision that one occurs during the last 30 d of the cy cle ; (d) one shutdown and one drain of one-half of the reactor coolant volume for m ajor maintenance followed by a restart, and occurring after the third hot shutdown and restart; (e) one cold shutdown at an operating boron concentration not less than 100 ppm; and (f) one refuelling shutdown.

The volumes o f liquid waste generated by normal operation and by the postulated shutdowns are shown in Fig. 8.

To accommodate the volumes of liquid waste that could be generated by this shutdown schedule the holdup tanks are sized to hold four com plete system volum es o f liquid w astes at 120°F.

The reactor coolant radioactivity used for the design of the system was based on the nominal 1% defective-fuel cladding. In addition to the fission products and corrosion products, the coolant contains boric acid and from 1 to 11 ppm of potassium ions from the lithium hydroxide used as pH control chem ical to minimize corrosion . Hydrazine may also be used at startup as an oxygen scavenger whenever the coolant has been exposed to air.

When the coolant boron concentration is to be reduced or increased, demineralized water or boric acid solution, respectively, are added in the chemical and volume control system of the reactor- coolant loop. The liquid level in the volume control tank rises, and coolant norm ally flowing into the tank is diverted into the radioactive- w aste-d isposa l system at a flow -rate of either 45 o r 90 ga l/m in . The coolant waste is first sprayed into the vapour space of the flash tank in which a nitrogen blanket is maintained just above atmospheric pressure. Hydrogen contained in the coolant flashes and diffuses, carrying with it an estimated 50% o f the fission gases. From the flash tank the liquid is pumped to one of three 7000-ft3 holdup tanks.

From the holdup tanks the liquid is pumped through mixed-bed dem ineralizers which can be used either individually or in series. Two of the ion exchangers in series are used at the start of operations until effluent samples from the first indicate a gradual rise in activity. At this time the third ion exchange unit is placed in series operation downstream and, when needed, a fourth unit is added. When the fourth unit is placed in operation, the first unit is removed and its resin flow flushed to the spent resin storage tank and recharged.

2 0 7


This unit is the first "chem ical shim" plant to rely on ion exchange for collection of all radioactive nuclides. In the analysis of the design, 137Cs, 131I, "M o , 133I and 134Cs are considered to be the most significant. Caesium breakthrough is determined prim arily by the lithium added for pH control. Resin lifetim e was calculated on the basis of 6 ppm of lithium instead of the expected concentrations of 0.2 to 2. 1 ppm. Iodine breakthrough is not expected during any core cycle.

From the dem ineralizers the liquid waste flows to a liquid gas stripper in which the remaining gaseous nuclides are removed. The liquid then is pumped to one of two monitor tanks. After filling, the tank contents are sampled and analysed, and the contents pumped to the plant circulating-w ater effluent line through a cartridge filter, an activity monitor and alarm, and an automatic shutoff valve actuated by high-activity.

Potentially dirty liquids containing detergents, dirt, oil, floor drainings, chem icals, e tc ., are transferred to the decontamination drain tank for filtration before discharge. If the activity is sufficiently high these can be transferred to the,holdup tanks for processing the demineralizer. Samples of coolant and process liquids with sufficient activity and with no other contamination are drained to the radiological laboratory drain tank whose contents are periodically pumped to the holdup tanks for processing . Other samples which are contaminated with reagents, and the detergents and wash water from cleaning operations, are drained to the decontamination drain tank. Leakage of coolant or other liquids at pump seals, valve stems, e t c . , are collected in sumps which are automatically pumped to the decontamination drain tank for discharge or processing.

Effluent liquids from general washrooms, showers and toilets are not collected , sampled o r m onitored. The effluent from the potentially radioactive shower is drained to the decontamination drain tank.

3 .2 .3 .3 .2 . Experience

Liquid-waste volume and activity releases are summarized by month in Table XXVIII. Up to the end o f 1967, the liquid radwaste system released a total activity of 0. 317 Ci with an average concentration in the circulating-water discharge of 8 X 10 '10 /nCi/ml.

2 0 8


TABLE XXVIII. RADIOACTIVE GASES AND LIQUIDS DISCHARGED AT SAN ONOFRE3

Period

Gaseous releases Liquid releases

V olume (ft3)

Activity(Ci)

Volume(gal)

Activity(Ci)

June 1967 2 236 0.0015 166 900 l.o x io"2Jul. 1967 0 - 98 660 6 .5 X 1 0 '4

Aug. 1967 4 722 0 .002 96 250 5 .5 X 1 0 '3

Sep. 1967 7 761 0 .9 6 48 555 8 .5 X 1 0 '3

Oct. 1967 11 748 1 .43 123 275 3 .1 X 10"2

Nov. 1967 8 712 1.07 72 464 4 .0 X 1 0 '3

Dec. 1967 3 052 0 .4 4 151 640 2 .6 X 1 0 '1

Compiled from Monthly Operation Reports for San Onofre Nuclear Generating Station.


3 .2 .3 .4 .1 . Sources

The m ajor quantity of gaseous radioactive waste originated with the liquid generated either at shutdowns or by the continuous dilution of boron contained in the reactor coolant during the core cycle. The concentration o f fission gases in prim ary coolant is expected to be approximately 230 pC i/m l with 1% defective fuel. Most of the waste gas volume consists of hydrogen and nitrogen. The coolant contains about 45 cm 3 (STP) o f hydrogen gas per kilogram o f coolant. The volume of xenon and krypton gases in one system volume of coolant would be less than 1 ft3, com pared with 300 ft3(STP) o f hydrogen. Nitrogen is used as a cover gas in the flash tank and waste tanks.

It is estimated that about 50% of the volatile activity w ill be removed from the coolant in the flash tank. These gases flow continuously to a gas surge tank.

2 0 9


The gas surge tank receives gaseous activity from the flash tank and gas stripper, as well as from the vent header within the containment sphere during normal operation. As waste-gas pressure rises in the gas surge tank, the w aste-gas com pressor starts and com presses the gas into one of the decay drums where it can be stored for decay before release to the vent stack. Before reaching the vent- stack fans the gas is passed through a roughing pre-filter and a high- efficiency filter.

3. 2. 3. 4. 2. Operating experience

Gaseous releases are tabulated by month in Table XXVIII. Up to 31 December 1967, gaseous releases had totalled 3. 90 Ci with an average concentration in the stack o f 1.2 X 10”8 juCi/cm 3, without com pression and decay storage. This can be com pared with the annual average licence lim it o f 2. 86 X 10~3 (iC i/cm 3.

TABLE X X IX . PLANT CHARACTERISTICS THAT AFFECT RADIOACTIVE W ASTE GENERATION AT CONNECTICUT YANKEE

Coolant Light water

Coolant cycle Pressurized water - chemical shim

Type of fuel Type-304 stainless-steel-clad U02(16 Zircaloy-4 clad fuel rods for testing)

Fuel failures 1% design

Control rod material Stainless-steel-clad Ag-In-Cd alloy

Other material exposed to reactor coolant

Austenitic stainless steels, Inconel

Coolant purification 3 mixed-bed demineralizers (45 ft each,non-regenerative), disposable cotton- cellulose filters

Additives to reactor coolant Hydrogen, boric acid, (possibly)ammonium hydroxide, hydrazine (after shutdown)

210


3 .2 .4 .1 . General

The Connecticut Yankee nuclear generating plant is located near the town of Haddam on the east bank of the Connecticut R iver about 22 m iles south of Hartford. The reactor is a Westinghouse pressurized-w ater type designed to produce, initially, 1473 MW o f heat and 490 MW gross e lectr ica l pow er. The turbine generator plant is designed to permit generation of 590 MW gross if the nuclear- steam- supply system should have adequate margin for such operation.

The reactor first achieved criticality on 24 July 1967, and the turbine generator was phased in on 7 August 1967. Up to 31 March1968, it had produced 1 157 300 g ro ss MWh o f e le c tr ic ity [2 6 ].

The plant characteristics that affect the nature of the radioactive waste are summarized in Table XXIX. The initial fuel core is clad with type 304 stainless steel with the exception of 16 rods which are clad with Z irca loy-4 for testing purposes. Figure 9 is a simplified flowsheet showing the sources, treatment and disposition o f radioactive waste at Connecticut Yankee. The system fo r Connecticut Yankee differs from those previously described by its treatment of boron recovery for re-use as part of the radioactive-waste processing line. The processes and equipment are summarized in Table XXX.

As in other pressurized-w ater reactors, the reactor coolant is purified by passage through a dem ineralizer and filter in a bypass loop. Lithium hydroxide is used for pH control. Each of the three purification dem ineralizers contains 45 ft3 o f nuclear grade resin. These are not regenerated but sluiced directly to a shipping container when exhausted. The filters employed are composed of a disposable cotton-cellulose fibre medium. Similar filters and dem ineralizers are employed for clarification of the spent-fuel-pit water. Secondary- w ater-so lids control is maintained by blowdown from each steam generator. If the blowdown becomes contaminated these liquids may be diverted to the waste-disposal system for processing.

. Connecticut Yankee operates under an AEC License, No. DPR-14. Under this licence, operating restrictions on discharges, have been established for gaseous and liquid wastes. The annual average r e lease rate of gaseous radioactivity discharged at the plant stack is limited to 10 000 /^Ci/s of dose equivalent 133Xe. The maximum re lease rate averaged over any one hour is restricted to ten times the

3. 2. 4 . _ C onnecticut Yankee

211


TABLE X X X . BORON RECOVERY AND W ASTE DISPOSAL SYSTEM EQUIPMENT CAPACITIES AND RATINGS (27): CONNECTICUT YANKEE ATOMIC POWER COMPANY

Liquid waste systems:

1 Primary drain collecting tank

2 Aerated drain tanks

2 Boron waste storage tanks

2 Test tanks

2 Primary drain tank pumps

2 Aerated drain tank pumps

2 Waste liquid transfer pumps

2 Test tank pumps

1 Aerated drains evaporator .

7 500 gal

2 500 gal each

75 000 gal each

16 000 gal each

200 gal/min each

50 gal/min each

40 gal/min each

40 gal/m in each

50 gal/h

1 Waste liquid transfer filter

1 First-stage evaporator bottoms pump

1 Second- stage evaporator bottoms pump

1 Evaporator recycle pump

2 Distillate pumps

1 Distillate accumulator

25 (im

555 gal/min

555 gal/min

8 gal/min

28 gal/min each

480 gal

Waste gas system:

1 W aste gas surge sphere 20 000 ft3

2 Waste gas blowers 10 ft3/min each

annual average lim it. The average release rate is based on a m eteorological dilution factor of 1000 from the plant stack, two-fan dilution flow of 3. 3 X 107 crrP/s and the MPC for 133Xe in unrestricted areas as specified in Title 10, Code o f Federal Regulations, Part 20. Liquid wastes are restricted to concentrations in the circu lating water discharge as established in T itle 10, Code o f Federal Regulations, Part 20.

Cost data for the waste-disposal system are not yet available.

212



3 .2 .4 .2 .1 . Sources

The solid-waste disposal system handles com pressible and non - com pressible waste material, evaporator bottoms, spent resin and expended filter cartridges.

Contaminated com pressible materials are baled in a size to fit a 55-gal drum. Unfilled spaces in the drum can be filled with concrete to provide shielding if required or with low -activ ity non- com pressib le wastes such as sm all tools.

Evaporator bottoms, s lu rries from the b oron -recov ery evaporators and the aerated-drains evaporator are mixed with cement powder in a 55-gal drum. The drum is rolled to provide hom ogeneous cement mix and after curing is transported to the yard s to r age area for ultimate d isposal off site by a licensed contractor.

Spent resin is sluiced directly to one of two 170-ft3 steel shipping containers placed in rein forced-concrete shipping casks. One of the shipping casks will normally be in place whereas the other cask may be in shipment. Spent resin from all radioactive-process de- m ineralizers and ion exchangers is handled in the sam e manner. The resin is flushed to the spent-resin shipping container using prim ary-grade water. When all spent resin has been transferred, the flushing flow is terminated and water in the shipping container is pumped out. The shipping container is stored in its casks in the yard storage area until it can be loaded onto a truck for disposal off site. The spent resin in its container is buried and the concrete cask is returned to the plant for subsequent re-use. Two shipments per year are expected.

The reactor-coolan t letdown filter, spent-fuel pit filte r and waste-liquid transfer filters are rem oved from serv ice when the pressure drop becom es excessive or the radiation level exceeds a pre-determined maximum,. The expended filter cartridges in their basket are ra ised into a filter rem oval shield and m oved by yard crane to a filter shipping container and a rein forced-concrete shipping cask. A special cask cover is positioned so that expended filter cartridges can be loaded in one of six positions form ed by dividing plates in the shipping container. When six expended filter cartrid ges have been placed in the container, a steel cover is positioned and a concrete cask shipping cov er is installed. The disposable

2 1 3


filter container in its shipping cask is then ready for o ff-s ite d is posal3.

M iscellaneous contaminated m aterials are placed in 55-gal drums, concreted for shielding, which are then stored for disposal off s ite .


No solid wastes had been shipped fo r o ff-s ite d isposal up to March 1968.

3. 2 .4 .3 . Liquid wastes

3. 2. 4 .3 .1 . Sources

Reactor-coolant waste enters the liquid radwaste from the prim ary drain collecting tank from which it is pumped to the boron-waste storage tanks. Liquid to be processed is pumped from the boron- waste storage tank through the hydrogen-cycle ion exchanger and waste-liquid transfer filter to the boron-recovery evaporators. The ion exchanger uses a hydrogen form cation resin . This resin r e moves 137Cs, which represents the m ajor fission product in the r e actor coolant, and yttrium, strontium and molybdenum from the liquid before its entry to the evaporator.

The evaporator is a two-effect system designed to operate normally as two units in series processing 20 gal/m in or as two units in parallel processing 40 gal/m in. Each evaporator stage contains a cyclone separator and entrainment elim inator, and is expected to provide reduction in boron concentration in the overhead flow by at least a factor of 200 from the liquid. The steam flows through a condenser into a distillate accumulator from which it is pumped to the w aste-disposal test tanks. Non-condensable gases present in the feed stream are rem oved both at the condensate outlet from the second-stage evaporator reboiler and at the distillate accumulator vapour space. These gases are discharged to the gaseous waste system.

When the concentration of boric acid in the first-stage evaporator bottoms reaches the recovery value, it is cooled toa temperature adequate to prevent crystallization in the pipe line, and pumped to

Recent information [26] indicated that the filters are to be shipped in 55-gal drums.

2 1 4


the boric acid tank. Whei} packaging of the evaporator bottoms is desired, the liquid is passed through the boric acid recovery cooler and transferred to the packaging station.

The condensate from the b oron -recov ery system is sampled and pumped either to the r iver via the se rv ice -w a te r -sy s te m - discharge header or to the primary-water storage tank. If additional processing is required it can be returned to the boron-recovery system for further treatment. All piping in the boron-recovery system is typp 304 stainless-steel, and liquid lines containing concentrated boric acid solutions (equal to or greater than 1% by weight of boric acid) are electrically heat traced.

Aerated-liquid wastes flow into the liquid-waste disposal system from the vent and drain system, and are collected in aerated-drain tanks. The liquid in these tanks is sampled to determine whether further treatment is required before final disposal. If further treatment is not required the liquid is pumped'into the river via the service-w ater-d ischarge header. If further treatment is required the liquid is pumped to the aerated drains evaporator.

The aerated-drains evaporator is a com m ercially available im m ersed tube-bundle-type evaporator capable of handling 50-ga l/h . The evaporator provides a decontamination factor of 104 to 107 with liquid-effluent activities of the order of 10"8 to 10‘9 AlCi/ml. The processed liquid is collected in a distillate tank and after sampling is passed to the river via the serv ice-w ater-d isch arge header. Aerated-drains-evaporator bottoms are pumped to the solid-w aste disposal system for packaging.

The total amount of liquid processed by the boron-recovery system is a function of plant-operating param eters and stage of core lifetim e. Size of the system is predicated on its ability to store and process liquid wastes produced by the following plant operations: (a) a refuelling startup for each core cycle; (b) a cold shutdown and restart immediately following initial full power operation of each core cy cle ; (c) one 50% 60-h load reduction per week; (d) one hot shutdown of at least 60-h duration every four weeks; (e) one cold shutdown at end of full power life and one at refuelling; (f) core stretchout of three-months duration at zero boron concentration. It is also necessary for the boron -recovery system to be capable of handling the dilution from two sequential hot shutdowns of at least 60 h at maximum power between each shutdown occurring at any time over the core life.

2 1 5


TABLE X XX I. RADIOACTIVE GASEOUS AND LIQUID WASTE DISCHARGES AT CONNECTICUT YANKEE a

Period

Liquid waste discharged Gaseous waste releases Grosselectricitygenerated

(MWh)Volume(gal)b

Gross . activity (mCi)0

Tritium(Ci)

Volume(ft3)

Activity(mCi)

Aug. and Sep.1967 -- — -- - -- 3 840

October 1967 283 050 59.15 1.83 1 100 9.1 56 340

November 1967 298 750 70.83 123.42 2 000 11.4 150 630

December 1967 162 500 124.72 95.31 - None 350 700

January 1968 181 100 130.95 215.19 8 100 624.8 240 210

February 1968 56 400 47.12 59.67 - None 335 670

March 1968 167 810 356.27 335.43 4 204 000 d 36.5 19 950

a Information received in personal correspondence with D.E. Vandenburgh, Connecticut Yankee Atomic Power Company [26]. k Before dilution (372 000 gal/min circulating water). c Excluding tritium, d Reactor containment purge.


The approximate volum e of liquid waste entering the boron - recovery system under various conditions is as follows; (a) dilution for cold shutdown to full power at the beginning of core life is equal to 1600 ft3; (b) for dilution from cold shutdown to full pow er just before the end of life the volume is 19 300 f t3.

3, 2. 4. 3. 2. Experience

Waste releases from the Connecticut Yankee plant are summarized in Table XXXI [26]. As indicated, the non-tritium liquid r e leases have been quite sm all and well within the design estim ates. Tritium releases, although they seem relatively high in comparison with the other nuclides, averaged about 2 X10'7 AiCi/ml in the 372 000 gal/ min circulating-w ater discharge or over four orders of magnitude below the 'MPC for this isotope.


3 .2 .4 . 4 .1 .- Sources

Gaseous wastes enter the system from the vent and drain system, as off-gases from the boron -recovery system and as gases evolved from liquid in the prim ary-drain collecting tank and boron-w aste storage tanks. These wastes consist alm ost entirely of hydrogen and the fission product gases, although nitrogen used fo r purging the air from the system may also be present in the gaseous wastes.

The gas piping is arranged as a cascade system . A ll gases enter the system via the w aste-gas header near the prim ary-drain collecting tank, and flow to the vapour space of one of the boron - waste storage tanks. The gases then flow to the vapour space of the second boron-w aste storage tank and finally to the w aste-gas surge sphere.

The gas sphere has a polyurethane-coated nylon cloth diaphragm attached at its equator. The sphere receives the gas displaced from the waste-liquid tanks upon filling and restores the gas to the'tanks upon emptying. O ff-gases continuously accumulate in the waste- gas surge sphere from plant operation and the operation of the boron- recovery system . When a quantity of o ff-ga s has accumulated in the waste-gas surge sphere it is sampled to determine its activity. The o ff-gas is then gradually drawn from the sphere by one of the waste-gas blowers and discharged through a fib re -g lass decay gas

2 1 7


filter by the suction of two 35 000 f t 3/m in ventilation and purge fans. An interlock prevents discharge of the waste gas while an insufficient ventilation discharge flow exists. The diluted mixture is d ischarged to the atmosphere via a 175-ft prim ary vent stack.

3. 2. 4. 4. 2. Operating experience

Gaseous waste experience is summarized in Table XXXI. As with the liquid wastes, the gaseous-waste releases are well below the perm issib le lim it of 10 000 m C i/s, averaging 58 n C i /s in the 70 000 ft^ /m in vent exhaust.

3 .3 . Present developm ents in PWR w aste-system design

Present developments in radioactive-w aste system s fo r pressurized-w ater reactors generally follow the outlines indicated in the San Onofre and Connecticut Yankee designs. The basic d ifference lies in the decision as to whether or not boron recovery is necessary or desirable. By far the majority of the stations currently under construction employ waste evaporators.

In the solid-waste area there are also few if any changes from the earlier designs. Connecticut Yankee is unusual in the provision of re-usable shipping casks for the spent filters and spent resins. Most station designs use 55-gal drums with admixtures to provide solidification or im m obilization of the contained concentrates or resins.

PWR plants uniformly provide decay storage for gaseous wastes. In m ost plants operating to date it has not been n ecessary to p ro vide such storage to meet the applicable lim its. However, in line with the practice of minim izing o ff-s ite exposure to the maximum extent practicable, these storage facilities are provided and used when necessary.

4. DISCUSSION

4 .1 . General

Waste management at nuclear power stations in the United States of A m erica has been uniformly excellent, as evidenced by the data reported above. The leve ls of re lease from these plants of both

2 1 8


gaseous and liquid wastes have been generally less than 10% (and frequently less than 1%) of the established lim its. Even during periods of operation with defective fuel cladding, the lim its have not been exceeded, attesting to the conservatism in the design of the radioactive-w aste system s.

Detailed information on the performance of system components is generally not available. This is prim arily because operation of waste system s at nuclear power stations has generally not been of m ajor operational or econom ic importance. There has been little if any difficulty in operating within established lim its for rad ioactivity releases, and a considerable amount of latitude is provided in the system designs.

Evaporators have been provided at practically all stations except San Onofre, yet only at Yankee has this equipment been either needed or used. At Yankee evaporation is used prim arily because the plant management made a voluntary decision not to pollute Deerfield R iver with boric acid; the concominant reduction in the radioactivity release is a bonus but is only of secondary importance.

The design bases for these system s are, by and large, highly variable from plant to plant. These re flect not only the expected maximum level of activity in coolant system s during operation but also the mode of operation intended for the plant and the degree of operator attention desired by the plant management. Such factors as the plant staff size , the plant load cy cle (i. e. whether base- loaded or load-follow ing) as w ell as previous engineering firm or utility company experience with specific types of equipment play a m ajor role in the final design of the waste system s.

An excellent example of this type of design influence is obtained by com paring the Oyster Creek Unit No. 1 owned by Jersey Central Power and Light Company and the Nine Mile Point plant described above. Both of these plants are about 500 000-kW boiling-water r e actors built by the General E lectric Company with s im ilar if not identical core design. The flowsheet fo r the radwaste system s for Jersey Central is almost identical with that for Nine Mile Point (F ig. 4), with only m inor differences in cross-connections between parallel w aste-processin g lines. However, the component s izes , particularly in the tankage volume, differ significantly between the two stations. This com parison is indicated in Table XXXII which lists the major equipment items for both Oyster Creek and Nine Mile Point. Nine Mile Point, for example, has about times the capacity for filter sludge than does Oyster Creek; the floor-drain sample-

2 1 9


TABLE XXXII. RADWASTE EQUIPMENT SIZE COMPARISON

Item

No.

Oyster creek

aCapacity

Nine mile point

No. Capacity3

1. Centrifuge - 20 gal/min 1 20 gal/min

2. Concentrated waste tank 1' 5 000 gal 1 5 000 gal

3. Concentrated waste mixer 1 i yd3 1 i yd3

4. Drywell equipment drain sump 1 000 gal 2 000 gal

5. Drywell floor drain sump 1 1 000 gal 1 2 000 gal

6. Filter aid tank 1 500 gal 1 470 gal

7. Filter sludge storage tank 1 9 000 gal 23 000 gal

8. Floor drain collector tank 1 10 000 gal 1 10 000 gal

9. Floor drain filter 1 300 gal/min 1 300 gal/kiin

10. Floor drain sample tank 10 000 gal 20 000 gal

11. Laundry drain tank 1 2 000 gal 2 000 gal

12. Pre-coat tank 1 600 gal 1 560 gal

13. Reactor building equipment drain tank 1 5 000 gal 1 5 000 gal

14. Reactor building floor drain sump 1 1 000 gal 4 000 gal

15. Spent resin tank 3 000 gal 1 4 000 gal

16. Turbine building equipment drain tank 1 3 000 gal 1 1 000 gal

17. Turbine building floor drain sump 500 gal 4 800 gal

18. Waste collector tank 1 30 000 gal 1 25 000 gal

19. Waste collector filter 1 300 gal/min 1 300 gal/min

20. Waste concentrator 1 15 gal/min 1 12 gal/min

21. Waste demineralizer 1 300 gal/min 1 300 gal/min

22. Waste neutralizer tank 2 12 000 gal 1 15 000 gal

23. Waste sample tank 2 30 000 gal 2 50 000 gal

24. Waste surge tank 1 100 000 gal 1 50 000 gal

a Total capacity.

220


tank capacity is twice that of Oyster Creek; the turbine J building floor- drain sump has over nine times the capacity of that at Oyster Creek; and the waste-surge tank is only half the capacity of that provided at Oyster Creek. These differences largely reflect the difference in operating philosophy and station-load pattern expected by the Niagara Mohawk Power Company and the Jersey Central Power and Light Company, since the plants are otherwise largely identical. It is therefore almost impossible to develop a uniform opinion concerning an optimum design for nuclear plants in general; information relating to the specific use of the plant and its personnel by the power company management is required for optimum plant design.

4 .2 . Future problem areas

Finally, it would be appropriate to discuss two aspects of im portance in the design consideration of nuclear power station waste management. The first is concerned with tritium and its release from these stations; the second is concerned with the contamination of secondary water in pressurized-water reactors due to steam - generator-tube leaks.

Tritium has been identified in the primary coolant of all light- water power reactor plants (Table XXXIII). In addition to its fo rmation by activation of hydrogen and naturally occurring deuterium in the reactor coolant it is known to be produced by ternary fission of 235u ancj by neutron reactions with boron and lithium. In an informal review of tritium behaviour in reactors, Smith 128] concluded that, in the absence of boron and lithium, the tritium derives mainly from fission and that it appears in the coolant after having diffused through the fuel and the fuel cladding. It appears on the basis of the data in Table XXXIII that such diffusion is more pronounced with stainless-steel than with Zircaloy cladding, possibly due to the hydriding ("tritiding") of zirconium by the tritium.

Although there is no difference in tritium production from fis sion in reactors of similar capacity and fuel type it apparently reaches higher concentration in PWR than in BWR coolants. Probably the most reasonable explanation for this difference lies in the use of additives in the PWR coolants. As indicated in the sections above, both boron and lithium may be used in PWR coolants for reactivity control and pH control respectively. An analysis has been made of the PWR tritium production and release to the coolant for a particular plant [31]. The total tritium from the core, including

221


TABLE XXXIII. TRITIUM LEVELS DETECTED IN WATER OF VARIOUS REACTORS a

Plant Date Source TritiumfriC i/cm s)

Fuel cladding

Boiling water reactors:

D resden-1 7 /2 /6 4 Condensate 2 X lCT2 Z irca loy and stainless steel

1 0 /6 /6 5 Condensate 7 X 10"3 Zircaloy

2 0 /1 0 /6 5 Reactor water 3 X 10"s Z ircaloy

Humboldt Bay 1 6 /5 /6 5 Condensate 3 X 10‘ 2 Stainless steel

Big Rock Point 2 2 /9 /6 5 Reactor water 3 X 10"2 Stainless steel

SENN Italy 3 /5 /6 5

3 /5 /6 5

Condensate

Makeup water

5 X 10‘ 4

< 7 X 10"6

Z irca loy

VBWR ?/6 3 Reactor water 5 X lO "2 M ixed

Elk River b 7/65 Reactor water 6 X 10"2 Stainless steel

Pressurized- water reactorst

PM - 3A McMurdo Reactor water 0 .8 -0 .9 Stainless steel

Saxton Reactor water 0 .5 Stainless steel

SELNI Italy Reactor water 0 .7 5 Stainless steel

Y ankee Reactor water 2 -3 Stainless steel

Indian Point c 1965 Reactor water 0 .1 5 Stainless steel

a Data assembled by Smith [2 8 ].b Tritium le v e l in Elk River coolant reported by C am pbell et a l. [2 9 ], c Ref. [ 3 0 ] ,

222


that produced by ternary fission, and from the boron in the poison rods during the initial core cycle, totals approximately 7000 C i/yr. Data from the Yankee and Saxton stainless-steel-clad fuelled reactors indicate that the fuel retains about 68% of the tritium produced. Assuming then that 30% of the core-produced tritium escapes, approximately 2100 C i/yr would be released to the coolant from this source. Tritium produced in the coolant by neutron reactions with 10B and 7Li indicate a total value of about 575 C i/y r from these sources.

Thus, for a stainless-steel-clad fuel the tritium produced by the activation of coolant additives would appear to contribute about 25% of the total tritium in the coolant. Data from the Shippingport reactor with Zircaloy—clad fuel (although at a much lower specific power) indicate that less than 1% of the ternary tritium is released to the coolant. This would indicate a much lower escape coefficient and resulting discharge with Zircaloy-clad fuel.

The major reason for concern with tritium is the inability to remove this nuclide from wastes and its relatively long half-life . Tritium has not been a waste-disposal problem at any power station primarily because of the relatively high allowable discharge limit in water of 3 X 10*3 **Ci/ml. It is possible to demonstrate that for any reasonable once-through condenser-cooling system the concentration of tritium in the circulating-water discharge cannot conceivably reach 1% of the permissible concentration for this nuclide. How^ ever, the condition may be considerably more stringent for those power plants which operate using cooling towers, and which therefore have a restricted supply of dilution water available. In such cases it is conceivable that the tritium concentration, particularly from a PWR, could approach the permissible concentration in the discharge from the plant before its dilution in the receiving stream. This is a factor which must be taken into consideration in the design and location of large nuclear stations on "d ry " sites. An alternative may be the evaporation of all tritiated water with exhaust to the atmosphere.

As a final consideration it is of significance to note that steam- generator leaks have occurred in most, if not all, operating pressurized-water reactors. The introduction of radioactive m aterials into the secondary water would create the requirement for decontamination of this system for which provision is not normally made in PWR design. Although the technical specifications for nuclear plants restrict releases to the environment to those spe

223


cified in the appropriate regulations, the inability of power stations to deal effectively with these contaminated liquids may create an economic hindrance by requiring that the station be shutdown because of an otherwise relatively minor leak. Of the plants examined only San Onofre and Connecticut Yankee provide capability for the draining of steam-generator blowdown to the liquid radwaste system. This aspect is being considered, however, in other plants and may well be included in future designs.

R E F E R E N C E S

[1 ] BLOMEKE, J. O ., HARRINGTON, F. E ., Management o f Radioactive Wastes at Nuclear Power Stations, Rep. ORNL-4070 (1968). .

[2 ] Radioactive Waste Handling in the Nuclear Power Industry, Edison Electric Institute, New York, March (1960).

[3 ] ELLIOTT, V. A . , MAXON, R .D ., NIXON, V .D ., MERRYMAN, J, W . T h e Dresden Nuclear Power Station” , 2n^ Int. Conf. Peaceful Uses atom. Energy (Proc. Conf. Geneva, 1958) _8, UN, Geneva (1958) 508.

[4 ] The Dresden Nuclear Power Station, Pwr React. Technol. 4 1 (1961) 56.[ 5] STONE, V .L . "Operating experience with boiling water reactors, " 3rd Int. Conf. peaceful

Uses atom. Energy (Proc. Conf. Geneva, 1963)5, UN, Geneva (1965) 209.[6 ] KIEDAISCH, W .I ., Commonwealth Edison Company, Private communication (1 April 1968).[7 ] KIEDAISCH, W. I . , Liquid radioactive waste handling, Chem. Engng Prog. 58 1 (1962) 79.[8 ] BRUTSCHY, F .J ., GILBERT, R .S ., OSBORNE, R .N ., "The behaviour o f corrosion products

in boiling water reactors", Corrosion o f Reactor Materials (Proc. Conf. Salzburg, 1962)i . IAEA, Vienna (1962) 133.

[9 ] MITCHELL, W ., "Humboldt Bay and Big Rock Point", Pwr React. Technol. _7 1, (1963-64)70.[10 ] Nucleonics Wk, _9 3 (1968).[1 1 ] HAUSLER, L .M ., HAUETER, R .L ., CHRISTIANSON, R. C . , "Big Rock Point nuclear plant

in-vessel modification o f irradiated reactor internals", ANS Conference on Reactor Operating Experience, Jackson Lake Lodge, Grand Teton National Park, Wyoming (1965).

[12 ] Docket No. 50-155, Final Hazards Summary Report for Big Rock Point Plant, 1, 2 and Amendments (1961).

[1 3 ] CARROLL, J. C . , SCHUYLER, J. O . , Humboldt Bay Reactor operating experience", Nucl. Saf. _6 4, (1965) 441.

[1 4 ] Report on the operation o f Humboldt Bay Power Plant Unit No. 3, covering the period February 16, 1967 through August 15, 1967, Pacific Gas and Electric Company (Sep. 1967).

[153 CARROLL, J .C ., Pacific Gas and Electric Company, Private communication (27 March 1968).[1 6 ] Docket 50-220, Facility Description and Analysis Report, Nine M ile Point Nuclear Power

Stations, as amended through March, 1968.[1 7 ] MILNE, G .R ., STOLLER, S .M ., WARD, F .R ., "The Consolidated Edison Company o f New

York Nuclear Electric Generating Station", 2nd Int. Conf. peaceful Uses atom. Energy (Proc. Conf.. Geneva, 1958) _8, UN, Geneva (1958) 483.

[1 8 ] Consolidated Edison Nuclear Steam Generating Station, Pwr React. Technol. 6_ 3, (1963)20.

224


[1 9 ] Indian Point Station, Semiannual Operations Report 1 to 10 (2 August 1962 - 31 December 1967), Consolidated Edison Company o f New York, Inc.

[2 0 ] SHOUPP, W .E ., COE, R .J .„ WOODMAN, W .C ., "The Yankee A tom ic Electric Plant", 2nd Int. Conf. peaceful Uses atom . Energy, (Proc. Conf. Geneva, 1958) 8, UN, Geneva (1958) 492.

[2 1 ] Docket 50-29, Yankee Hazard Evaluation Report, YAEC-167, V o lsJ .a n d 2 .[2 2 ] OperationReports49,61,73, and85, Yankee Atom ic Electric Company, Boston, Massachusetts

(February 1965, February 1966, February 1967 and February 1968).[2 3 ] BERNSEE, J.W . "Radioactive waste disposal at a nuclear power plant", 43rd Annual

Massachusetts Safety Council, Boston, Mass. # 15 May 1964.[2 4 ] Docket 50-206, Final Facility Description and Safety Analysis Report, San Onofre Nuclear

Generating Station, Southern California Edison Company, as amended through March 1968.[2 5 ] Monthly Operation Reports, San Onofre Nuclear Generating Station, Southern California

Edison Company![2 6 ] VANDENBURGH, D .E ., Connecticut.Yankee Atomic Power Company, Private communication

(2 April 1968).[2 7 ] Docket 50-213, Final Facility Description and Safety Analysis Report, Connecticut Yankee

A tom ic Power Company, as amended through March 1968.[2 8 ] SMITH, J .M ., Jr., Technical Notes on Tritium in Water Reactors, A tom ic Power Equipment

Department, General Electric Company, San Jose,. Calif. (1965).[2 9 ] CAMPBELL, R. J ., HALL, D ., KETTNER, D ., Elk River Reactor operating experience, Nucl.

Saf. 7 4 (1966).[3 0 ] Indian Point Station, Semiannual Operations Report N o .6, February 1, 1965 -Ju ly31, 1965,

Consolidated Edison Company o f New York, Inc.[3 1 ] Docket 50-282, Prairie Island Nuclear Generating PI ant, Facility Description and Safety

Analysis Report, Sect. 1 1 .4 (March 1968),

225


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FIGURES FOR APPENDIX 4


SOURCES(MAXIMUM DESIGN ES TIM A TE S )

TREATMENT

G A S E O U S W ASTES

MAIN CONDENSER AIR EJECTOR40*/.H2 , 20V. 02,36*/. AIRTURBINE GLAND SEAL EXHAUST SYSTEM 10B «cf/yr

RADIATION MONITOR C LO SES„ 7 . . v a lv e 'i f ' io 7 “u c / m c 'i s 'Y x " ( ? * / '2 x10 » c f / y r fl_ _ 1 ______________ -^ X S X T '_________________________J M ] ________ 5 5 __________ ~ _______

_____ I1 130' OF 3 0 " -DlA. PIPE ~ 20 min HOLD-UP TIME 2 ABSOLUTE . . . .

(AIR WITH 0.1*/. PROCESS G A S )CONTAINMENT VESSEL EXHAUST 2 xIO9 scf/vr(A IR )

LIQUID W ASTES

130 OF 30 ' - DIA. PIPE

130' OF 24“ - DIA.PIPE ~ 2 min HOLD-UP TIME _______________________40<^cf m_

BUILDING VENTILATION AIR

D IS P O S IT IO N (O P ER A TIO N A L EX P ER IEN C E)

TO ATM OSPHERE

REACTOR ENCLOSURE PIPING 140000 GAL/VrbAND EQUIPMENT DRAINS 10- ® |ic/cc 5000 GAL EA. -tx>

REACTOR ENCLOSURE FLOOR DRAINS TURBINE BUILDING FLOOR AND I 10~3 EQUIPMENT DRAINS

RESIN SLUICING W ATER

100000 GAL./ vr

|ic /cc

ONE CARBON-STEEL DRAIN TANK

ONE 200gpm REACTOR FILTER ONE 200 gpm TWO ALUMINIUM

M IXED-BED WASTE WASTE H O LD -U P DEMINERALIZER TANKS

TO C LEA N WATER 9x10 6 G AL/yr

S TOR AGE FOR REUSE <KT*nc/cc

<3x1 04 pc /scc

<100_>*£./NEGLIGIBLE

"ACTIVITY

44000_scfm_NEGLIGIBLE A C TIV ITY

7x 10 j*c/*«c LICENSED ANNUAL AVERAGE RELEASE R A TE 0 ; MAXIMUM ANNUAL. 25 000 p c/ scc

300' UN LINED CONCRETE STACK

5000 GAL EA.

40000 GAL./ yr

<10'® pc/cc SECONDARY STEAM GENERATOR 10~® GAL /vr

3 x10-* p.c/cc

TWO CARB O N - S TE E L REACT. & TURBINE BLDG.

t FLOOR DRAIN TK S

25 000 GAL

ONE CARBON-STEEL WASTE COLLECTOR TANK

ONE 200 gpm WASTE FILTER

25 000 GAL EA

BLOW -DOW N

LAUNDRY W ASTES 20000 G A L/yr

<10-4 pc/cc

LABORATORY WASTES

SHOP FLOOR DRAINS

20000 O AL/yr <10*4 (ic /cc

40000 GAL/yr 1 0 " 4 j i c / c c 10000 GAL/yr 10“ 2 lie /cc

CONDENSATE DEMINERALIZER 160000 GAL/vr

EQUIPMENT DECON.& FLUSHING

REGENERATOR F ILTER BACKW ASHES

1 f ic/cc

F U E L CANAL W ATER 440000 GAL /yr

S O L ID W ASTES

SPENT R ESINS

1 ^ |jc /cc

900 f r /yr

ONE ALUMINIUM WASTE H OLD -U P TANK

CONDENSER COOLING WATER

OVERFLOW

TWO 30-in C EN TR IFU G ES

2 x 10® GAL /yr3 c /yr

167 000 gpm TO ILLINOIS RIVER

ONE SLUD6 E RECEIVER TANK

AT LICENCED ACTIVITY R ELEASE RATES IT IS ESTIM ATED THAT DOSE RATE TO O F F -S IT E PERSONNEL WILL NOT EXCEED MAXIMUM PERMISSIBLE RATES OF 0.5 rem PER YEARMOST OF THE ACTIVITY IS FROM LEA K A G E DURING S C R A M S -A B O U T 1000 GALLONS PER YEAR , CONTAINING ONE (ic/cc. DAILY LEA K A G E IS LESS TH AN 10"® (ic /c c

FUEL POOL FILTER

TO POOL

. % 250000GAL

)0 opmSTORAGE.

FILTERS

4000 c / y r

4000 ft3 /vr

M ISCELLANEOUSNO E S T IMATE

OUTDOOR WASTE STORAGE TANK PAINTED

ONE UNDERGROUND CARB. STEEL CONCENTRATED WASTE STORAGE TANK

2300 ft3TEMPORARYSTORAGE

50000GAL

BALER

ONE UNDERGROUND STAINLESS STEEL SPENT RESIN STORAGE TANK

b ##3 ONE STORAGE BUILDING2000 ft

*Tl200<t3TWO UNDERGROUND CONCRETE STORAGE VAULTS

(RM) RADIATION MONITOR

AUTOM ATIC VALVE

MECHANICAL VALVE

INTERIMO N -S IT ESTORAGE

IX]set STANDARD CUBIC FE E Tscfm scf PER MINUTEgpm GALLONS PER MINUTE c - CURIESuc/cc MICRO CURIES PER

CUBIC CENTIM ETER |ic/s<c MICRO CURIES PER SECOND GAS FLOW LINES INSTRUMENT LINES LIQUID AND SOLID

FLOW LINES

F IG .l. Waste-management system at Dresden Unit N o .l .


S O U R C E S

(DESIGN ESTIM ATES)

T R E A T M E N T D I S P O S IT I O N(O P ER A TIO N A L E X P E R IE N C E )

G ASEOUS W ASTES

EQUIPMENT VENTS

RADIATION MONITOR C LOSESVALVE IF 10c/see IS E X C E E D E D FOR 15 min

MAIN CONDENSER AIR EJECTOR 24.5 cfmH2 ,0 2 , AIR, WATER VAPORS, ACTIVATED AND NOBLE GASES TURBINE GLAND S E A L EXHAUST________________________________ U S cfm

© ------- 1 ----- I

VENTILATION AIR REACTOR CONTAIN M ENT SPHEREAND PART TURBINE BUILDING

CHEMICAL LAB. AND COUNTING ROOM

LIQUID WASTES

MAXIMUM MAXIMUMEXPECTED EXPECTEDACTIVITY BATCH VOLUME

EMERGENCY COND. DRAIN OVERFLOW

10 •< u c / m l

10- 3 u c / m l

3000 UAL.

5000 GAL.

I I V U N B 3 T 3 IE M l/ H A in j In n iD I 1 UK «.

DECONTAMINATION FLUSHING FUEL CASK6 uc/ ml 500GAL.(50aDm) SHUTDOWN AND CLEAN UP SYSTEMS0 500 GAL. FMERGENCY POISON SYSTEM DRAIN

10*3 uc/ ml 5000 GAL. ENCL. FLOOR DRAINS AND WATER VALVE PIT1 uc/ ml 1500 GAL. COND. DEMIN. RESIN TRANSFER 8. BACKW ASH

lO-3 uc/m l 4000 GAL. RESIN REGEN. RINSE WATER 110" 3 uc/ml 100 GAL. FLOOR DRAINS -TU R B . BLDG.10 "3 uc/ml 200 GAL. F.W. H EATER S. CONDENSER DRAIN MAINT.10“ 3 uc/ml 100 GAL. R ELIEF VALVES10"3 uc/ml 13 qpm PUMP SEALS. MECH. VAC. PUMP SEAL WATER10'S uc/ml 5 GAL. STACK DRAIN10"3 uc/ml 10000 GAL. CONCENTRATOR DISTILLATE10“7 uc/ml 100 GAL. FLOOR AND EQUIPMENT DRAINS0 .0 2 -6 uc/ml 500 GAL. RESIN TR ANSFER TO CLEAN -U P SYSTEM0.02 - 6 uc/ml 310 GAL. C L E A N -U P SYSTEM SAMPLE & LEAKAGE DRAIN1 0 -4 -6 uc/ml 400 GAL./hr CONTROL ROD SYSTEM DRAINS10*3 uc/ml 4000 GAL. H YD R O -TES TS

------------m 7 n "H O L D U P T IM E

--------------- - r \ S \ j --------------- ------------.H? L.DL!.P - T1.M-E ____________________ FLO W F R O M L O W E S J .T O W A RD H IG H EST . PR_0BAB1LJT_Y_

- - + M - ----------A B S O LU TE FILTER

-90 see

OF CONTAMINATION

ABSO LU TE FILTER

ENCLOSUR E ^ DIRTY SUMP 500 GAL.

EXPECTED RATES 8i VOLUMES

R ATEgpm

NORMAL VOLUME GAL ./day

MAX.VOLUMEGAL./day

I-OO*

50 500-1500 7000100

TURB. BLDG. DIRTY SUMP 500 G A L .

500 - 1000 1000 GA L . ^100-1000 2500

0.02-6 uc/ml 2.500GAL.(5Qgpm) RECIRC. PIPING DRAINS10~4 uc/ml 10000 GAL. REFUELIN G TANK SYSTEM SW ELL0.02- 6 uc/ml 50 GAL./hr VALVE SEAL L E A K -O F F. INTERVALVE DRAINSK T^ uc/ml 380 GAL./day SAFETY VALVE DISCHARGE. STEAM LINE

6 uc/ml0.02 - 6 uc/ml

uc/ml~

50 gpm PUMP SEAL LEAKAGESAMPLE DRAINS

15 G AL. /hr STEAM DRUM SAM PLE DRAINS

10~3 Uc/mTlOOqpm MISC. RELIEF VALVE BLOW -OFF6000 GAL. INITIAL TE S T CONTROL RODS

- 22-50

RADW ASTE SUMP 100 GAL.

0-100050001000GAL. /mo

RECVR.5000G A L.

TANK (D IR TY)

blXh.

4xj*RECVR. TANK (D IR TY )5000 G A L.

-CXh

A T M O S P H E R E

ANNUAL AVER AG E R ELEASE RATE ■< 30000 fic/scc

I__5TWO 30000 cfm FANS

* . ROOF EXHAU ST

SOCK N FILTER

rtX}*

50 1000 - 3000 8 000

ip -5 uc/ml_______ 180 qpm______ FU EL STORAGE PIT0 .0 2 -6 uc/ml 4000 GAL. C L E A N -U P SYSTEM10~7 uc/ml 10000 GAL. EM ERGENCY COND. DRAINSft0 2 -6 uc/ml 16000 GAL. C L E A N -U P SYSTEM AND FU EL PIT1 uc / ml 120000 GAL. FLUSH. DECON. PRIMARY SYSTEM

180 10000

ENCLOSURE C LEAN SUMP 500 GAL.

7 uc/m l0.1 uc/ml 10-* (tc/ml

3000 GAL. REGEN. CHEM ICALS100 GAL. CONCENTRATOR RINSE100 GAL.

10~7 uc/ml 5 GAL10*4 uc/ml 5 G AL.

WASHIN6 MACHINE LAUNDRY SINKACCESS CONTROL AREA SINK

10-7 uc/m | 100 GAL. EMERGENCY SHOWER

-CXJ-RECVR.TANK (C L E A N ) 5000 GAL .

4 X J *RECVR.-TANK(C L E A N )5000 GAL.

CARTRIDGE FILTER

75gpm □T5EHI5T

ipmJHOLDTANK5666 GAL.

HOLDTA N K

50 3000

CHEM ICALADDITION

3000- tX H

LAUNDRYTANK

10~5 uc/ml 50 GAL. SHOP DECONTAMINATION

SOLID WASTES5 c l l t ? .________HTTP0 5 c / ft3

200 » 3 / y 120 U a /vr60 f»3/vr

C L E A N -U P DEMIN.RADW ASTE DEMIN.

/vr CONDENSATE DEMIN.

250GAL I t

10 200 300

CHEM. RECVR. TANK 0

M X H l

RATE 75 gpm NORMAL TOL. 2 -1 5 0 0 0 G A L./d a y

H X h

MAX. VOL.70000 G A L./d a y

5000 GAL

CONDENSATE

TANK 25 250 1000

roSo’ SiL

J

EVAPORATOR

TO CANAL .NORM AL- 0 TO 70000 GAL /day

MAX: TO 70000 GAL./day A C TIV ITY : 6x10-3 ,je/cc

5.8 CUR IES /yr CANAL FLOW : 50000 9pm

HX3- . TO CONDENSATE STORAGE NORMAL: 0 TO 15000 G AL ./d ay MAX.- TO 70000 G A L./d a y A C T IV IT Y ' < 10*3 (ic Icc

<?M>

f tIX

RADIATION MONITOR

AUTOM ATIC VALVE

SLUICE^ W ATEROVER FLOW

CONCENTRATE TORAGE TANK

M IS CELLA NEO U S W ASTESSP EN T RESIN TANK 10000 GA L .

500 JU Al1200ftS|

DRUMS V A U LT

< ------------ 1“' ‘BALER STORAGE. -

INTERIM O N SITE . STORAGE b

a RUBBER LINED C A R B O N -S T E E L TAN K

b SHIPMENT BY TRUCK TO COMMERCIAL BU R IA L GROUND

NON AUTOM ATIC VALVE

c CURIES|ic MICROCURIESgpm GALLONS PER MINUTES cfm CUBIC F E E T PER MINUTE----------GAS FLOW LINES INSTRUMENT LINES

______ LIQUID AND SOLIDFLOW LINES

FIG.2. Waste-management system at Big Rock Point.


SOU RC ES (OPERATIONAL EX P ER IEN C E)

TREATMENT D IS P O S IT I O N(O P ERATIONAL EX PERIENC E)

TO A T M O S P H E R E

GAS W A S T E S .MAIN CONDENSER AIR EJECTO R 18 cfm ~ 1 » 107 s c f/ y r( 207.H 2, t0V.O2 , 70'/ .A IR )

TURBINE GLAND SEAL EXHAUST SYSTEM 10B s c t/ vr(AIR P LUS ACTIVATION AND GASEOUS FISSION PRODUCTS)

DRYWELL PURGE b( AIR AND AR G O N - 4 ’ )

REFUELING BUILDING GAS TR E A TM E N T SYSTEM d

RADIATION MONITOR CLO SES VALVE IF 0.5 c / scc IS EXCEEDED

OAS TREATM ENT SYSTEM

LIQUID W ASTES

LAUNDRY AND HOT SHOWERS AO 000 GAL./yr< 10- 6 (ic I ml

STACK DRAIN & CONDENSATE TANK DRAIN 120000 GAL./yr

- - - - - - '-IB--' - -24“ - DIA. PIPE ~ 18 min HOLDUP TIM E ONE ABSOLUTE

-------------------------------------------------------------FIL-TE- -------------1 2 " - DIA. PIPE ~ 50 sec H OLD UP TIM E

18 S£f m _ _

200 scfmI--------------_ J

0 OR 2000 scfm

0 OR 134 scfm

<10“ ® uc/mlLABORATORY & TURBINE BLDG. FLOOR

DRAINCONDENSATE DEMIN.REGENERATION TANKS 400000 GAL./yr

TURBINE BLDG. CLOSED DRAINS

REFUELING BLDG. CLOSED & FLOORDRAINS

EMERGENCY CONDENSER DRAIN

< 5 x 10” 5 jic/cc

250GAL

ONE C A R B O N -S TEE L LAUNDRY WASTE TANK

-ai £ b >

VENTILATION AIR

— [X h

ONE CAUSTIC SCRUBBER 134 s c fm , 2.5 g p m , 5 NaOH

3000 GAL

REACTOR DRYWELL DRAIN 8i FLA N G E 40000 GAL./yrLE A K DETECTOR SCRAM SYSTEM

REACTOR CAISSON FLOOR & CLOSED DRAINS

REACTOR DRAIN

CAISSON PIPE CHASE FLOOR DRAINS

<0.2 |4c/ml

ONE TURBINE BLDG. DRAIN TANK

CAISSON SEEP AG E NEGLIGIBLE

<10~® p c/m l

500

ONE REACTOR EQUIPMENT DRAIN TANK

50 GAL."

SOLID WASTE

SPENT RESINS FROM CONDENSATE. CLEANUP ~ 3 5 tt3/yrAND RADWASTE DEMINERALIZER

FILTER CARTRIDGES

ONE REACTOR CAISSON SUMP TANK

230 GAL. 500<10-4

7500 G AL. EACH b

$

THREE RECEIVER

Y TANKS

* 00-

12000 scfm

MAXIMUM ANNUAL AVERAGE R ELEA SE R A TE ; <50000 ^ic /sec

250' UNLINED CONCRETE STACK

(RM)

TH R EE FILTER S 50 gpm CARTRIDGE TYPE ONE 50 gpm

MIXED BED RAD WASTE DEMINERALIZER

-CXJ-

GAL./day |i.c /ml '

ONE RADWASTE SUMP TANK

<

TO CONDENSATE

600 000 GAL./yrSAM PLINGSTATION

1.0 c / y r

-^ M )

io - j

STORAGE FOR REUSE

ONE 1-gpm CONCENTRATOR

>10"3 (ic/ml

10 gpm

©

scfscfmgpm

Kc

RADIATION MONITOR

AUTOM ATIC VALVE

LIQUID SAMPLER

CURIES LES S THAN MORE THANSTANDARD CUBIC FE E T scf PER MINUTE GALLONS PER M INUTE MICROCURIES

GAS FLOW LINES

INSTRUM ENT LINES

LIQUID AND SOLID FLOW LINES

5000GAL.

bEACH

TWO CONCENTRATED WASTE STORAGE TANKS

~100 ft /yr DRUMS

ONE SP EN T RESIN STO R AGE TANK

CONDENSER COOLING WATER

a DRYWELL PURGE USED ONLY WHEN ACCESS TO DR YW ELL AND R EAC TO R VESSEL IS REQUIRED

b P LA S TIC -C O A TE D C AR B O N -STEEL TANK

c SHIPM ENT BY TRUCK TO COMMERCIAL BURIAL GROUND

d THE BUILDING WOULD BE HELD UNDER A NEGATIVE PRESSURE WITH CONTAMINATED AIR EXHAUSTED THROUGH THIS SYSTEM IN TH E EVENT OF A R A D IO LOGICAL R E L E A S E ACCIDENT WITHIN THE BUILDING

TOHUMBOLDT BAY

M ISCELLAN EOUS WASTE -1000 ft3/yr '

1200 f»3 *

INTERIM ^ O N -S IT E -

STORAGE

ONE CONCRETE STORAGE VAULT

360 ft 2 c

ONE STORAGE BUILDING

O F F -S IT E DISPOSAL BY AEC LICENSED CONTRACTOR

FIG.3. Waste-management system for the Humboldt Bay Reactor.


S OUR CE T R E A T M E N T DI SPOSAL

GASEOUS WASTESMain Condenser Exhaust

Turbine Gland Seals

140 cfm

STEAM JE T AIR EJECTORS

30 Min. delay

1900 cfmMuch. Vacuum

Pumps

LIQUID WASTESLOW CONDUCT IV ITY

Reactor Bldg Equip Drains 4470 gpd @ 0.005 uCi/ml

Turbine Bldg Equip Drains- 52 16 0 gpd @ 10 -1 uCi/ml

Drywell Equip Drains 5330 gpd (a) 0.1 uCi/ml

Waste Bldg Equip & off Gas Drains-12,500 gpd (ffl 0.02 uCi/ml

Cleanup & Fuel Pool Systems HIGH CONDUCTIVITY; Floor Drains

ORA IN TANKS

Reactor Bldg2000 gpd @ 1 0 ~ 4 uCi/ml

Turbine Bldg2000 gpd @ 1 0 ~ 6 wCi/ml

Drywell Drains2 50 0 gpd @) 10- ^ uCi/ml

Waste Bldg.2000 gpd @ 1 0 ~ 4 uCi/ml

Sample & Lab5 0 0 gpd @ 1 0 “ 4 uCi/ml

Lab & Laundry Drains1500 gpd 4 * 1 0 ~ 5 uCi/ml

C H EM IC AL W ASTES

Regeneration Solutions 8700 gpd @ 2 * 1 0 - 3

Decontaminat ion500 gpd @ 1 uCi/ml

SOLID WASTES

C O M PR ESS IB LE WASTES

N O N -C O M PR E SS IB LE WASTES

(4000 50I)

(4800 gal)

(2000 * 1)

(3000 gal)

SUMP TANKS

LAUNDRY DRAIN TANKS(2000 gal)

CHEMADOITION

TANK

2 HEPA FILTERS

1.75 min. dtloy

-A A ^ 2040 cfm 0 590 uCi/sec

WASTE COLLECTION TANK

(25,000 f>l)Recycle

Filter Aid

WASTE SURGE TANK

(50.000 «al)

Iter Sludge

WASTE COLLECTOR FILTER

(300 gpm)

Filter Aid

FLOOR DRAIN COLLECTOR TANK

(10,000 gal)

TO WASTE

NEUTRALIZER - TANK

Turbine BldgWaate Dispoaal

9000 cfm170.000 cfm

SAMPLER

VENTILATION

EXHAUSTReactor Bldg

35,000-70.000 cfm

> « -

CONDENSATESTORAGE

a

WASTE DEMINERALIZER (300 gpm-150 ft3)

WASTE SAMPLE TANKS

(50,000 go I)

C Z D

FLOOR DRAIN SAMPLE TANK

(20,000 gal)

F ILTER SLUDGE STORAGE TANK

(23,000 gal)

SPENT RESIN STORAGE TANK

(4000 gal)

MOISTURE ABSORBENT

0-400 gpd I

| T iS C i7 5 i* WASTEMIXER

1610 ft3/yr 755 ft3/,

2500 ft /yr - 0.5 Ci/yr

WASTE NEUTRALIZER TANK

(15,000 gol)

WASTECONCENTRATOR

(12 gpm)

WASTE CONCENTRATOR

TANK[5000 gal)

lOOQftVyr - 10 Ci/yr

HYDRAULICBALER

5:1 500 ft3/yr

FLOOR DR. 250 Ci/yr COLL. TANK

()i yd3) (3®Q0 gpd « 0.001 uCi/ml)

DRUMS (55 gal)

120 Ci/yr

CENTRIFUGE(20 gpm)

WASTE HOPPER (57 yd3)

240 FO OT STACK

LEGEND

RM RADIATION MONITOR

CH CHECK VALVE

1 ROUGHING FILTER

1 HEPA

M NORMALLY OPEN VALVE

M NORMALLY CLOSED VALVE

CIRCULATING— ------- - WATER

<------- } (240,000 gpm)

SHIPMENT OFFSITE

FIG. 4. Waste flowsheet at Nine Mile Point Nuclear Power Station.


SOURCES

(D ESIG N ES TIM A TES )

GAS WASTESSW EEP GAS SYSTEM

PRIMARY COOLANT SYSTEM DEAERATOR

TANK VENTS

TRE AT M EN T DISPOSITION(OPERATIONAL EXPERIENCE)

CONTAINMENT VENTILATION AIR

LIQUID WASTES

REACTOR PLANT LIQUID WASTES

SYSTEM LEAK AG E ( NORMAL OPERATION ) 6000 GAL./yr

5 V> 000J e f / yr _ _ q _— *■ 0-01 uc / cc “ O

f 8 > W O WATERSEALED i1 COMPRESSORS 1

16 scfm EACH 1

iiclccONE E a RB -S T E E L I c » L ER GAS SURGE TANK | = > tA L tu

0 -10 psig

- r ~ Q - - - * nT I—J rrI TWO CATALYTIC

COMPRESSORS ' COMBINERS I 4 scfm EACH 1 4 « f n > EACHI

O X YG EN --------1

4 ficlccSAMPLING DRAINS (NORMAL OPERATION) 36000 GAL./yr

DRAINING EQUIPMENT

COOLANT EXR&NSION(STARTING UP)

K n c / c c

34000 GAL. I yr4 jic / cc

NEUTRON POISON-DILUTION t FLUSHING 300000 GAL./yr(STARTING UP )RESIN S LU IC E WATER

2 |jc/cc 2000 GAL./yr

CONTAINMENT AND OTHER4x10‘ 3 \iclcc 500 000 GAL, /yr

CONTAMINATED AREA DRAINS

CHEMICAL WASTESEQUIPMENT DECONTAMINATION

5x10~4 (ic/cc

20000 GAL./yr I0 "2 (ic/cc

PRIMARY LOOP DECONTAMINATION 34000 GAL./yr

NEUTRON SHIELD TANK WATER

3 (ic/cc

22 000 GAL./yr

S P E C IA L WASTES LABORATORY SINKS

1 He /cc

9000 GAL./yr

I

LIQUID WASTE COLLECTION SYSTEM INCLUDING NINE

STAINLESS STEEL

TANKS WITH

^ AGGREGATE)---------» -CAPACITY OF ABOUT 28000

GALLONS.SOME

WASTES GO DIRECTLY TO 75000 GALLON

TANKS

1000 G A L/yr

COOLER

_ _ £ ^ 2 5 0 f t 3___ g ____ j i j 500 f t 3 FILTER 35'scfm

ONE CARB -S TE E L FISSION GAS SURGE TA N K ,0 -5 p s ig

FOUR C A R B .-S TE E L PRESSORS. GAS HOLDUP TANKS lOscfm o - 6 0 d s i q EACH p 9l ““ 'I

ONE ABSOLUTE ER 35 s<

m - -

TO ATM OSPHERE

___ ^ | ^ | 0NE_Aj50L_. H LTE R ________

CONNECTION FOR F U T U R E _ G A S _ ____BOTTLING STATION

____ VENTJLATION_AIR__________________

SUPERH_E AT ER_FL_UE_ GAS_ ^ _ _ j

CONDENSER WATER E F F L U E N T

AVERAGE R ELEASE R A T E : 1 |ic /see

400ft STEEL FLU E GAS STACK

260 000 gp m

LAUNDRY DRAINS3x10"3 [ic/cc |

180 000 GAL./yr

WASTE HANDLING SERVICE

5x10"^ |ic/cc

14000 GAL./yr

75 000 GAL. EACH

ONE 100 gpm WASTE CIRCULATION PUMP

_^T\_I

A

ONE 12 gpm INCONEL GAS STRIPPER

ONE 12 gpmIN CON SlEVAPORATOR

-QFOUR SPHERICAL S TA IN L E S S STEEL LIQUID W ASTE COLLECTION TANKS

3000 G AL./yr

A *3350

GAL. EACH

BUILDING DRAINS 1.2 p.c/cc

NORMALLY NON-ACTIVE WASTESDECAY HEAT COOLING SYSTEM_________ 7000 GAL./yr

CEM ENT AND WATER

SOLID W ASTESSPENT RESINS 500 ftJ /yr

1400 cFILTE R S

1500GAL PaCTT

COMBUSTIBLES 100 ft /yr

NON - C O M BU STIBLESNO ESTIM ATE OF

FOUR STAIN LESS STEEL SPENT RESIN ST0RA6E TANKS

ONE 12 gpm DEMINERALIZER j w q STAIN

LE S S STEEL MONITORING TANKS

TWO S TA IN LE S S S T E E L EVAPORATOR C O N C EN TR A TE TANKS

/ \ _

20 c/yr

EN C ASEM EN T IN CONCRETE IN 55 GALLON DRUMS

LARGE EQUIPMENT STORED ON SITE

TO HUDSON RIVER

RADIATION MONITOR

AUTOM ATIC VALVE

STANDARD C UB IC F E E T scf PER M INUTE GALLONS PER MINUTE CURIESMICROCURIES PER CUBIC CEN TIM ETER

POUNDS PER SQUARE IN C H -G A U G E

GAS FLOW LINES

INSTR UM ENT LINES

LIQUID AND SOLID FLOW LINES

BURIAL4 c / y r

FIG. 5. Waste-management system at Indian Point Station.


SOURCES TR E A TM E N T

GAS W ASTESTO ATMOSPHERE

DISPOSITION( OPERATIONAL EXPERIENCE )

5 c / y rR ELEASE RATE

LIQUID WASTES150ft STEEL VENTILATION STACK WITH VENTURI AT TOP

DIAM ETER PIPE ABOUT 8'ABOVE TOP PRIMARY AUXILIARY BUILDING

850 000 G A L./yr 0.01 c / yr c 103 c/yr 3H

TO D E E R F IE L D RIVER

SOLID WASTESSPENT RESINS

aoo.PRUMs#^. 0 u r i a l

o THIS TANK RECEIVES WATER BLED FROM THE PRIMARY SYSTEM

b FOR THE DESIGN BASIS, R EFUELIN G SHIELD WATER IS NOT CONSIDERED A WASTE c EXCLUSIVE OF TRITIUM

d BURNING RATE IS 40 lb s /hr&

RADIATION MONITOR c CURIES|1C I c e MICRO CURIES PER

AUTOM ATIC VALVE CUBIC CENTIM ETER--------- GAS FLO W LINES

< LESS THAN INSTRUMENT LINES> MORE THAN LIQUID AND SOLID

APPROXIMATELY FLOW LINES

s c f STANDARD CUBIC F E E Ts c fm s c f PER MINUTEgpm GALLONS PER MINUTE

FIG.6. Waste-management system at Yankee.


G A S E O U S W A S T E S

FIG. 7. Waste flowsheet at San Onofre Nuclear Power Station.


BORO

N C

ON

CEN

TRA

TIO

N

(PPM

)

S Y S T E M V O L U M E S OF L I QUI D R A D W A S T E S

FIG. 8. Liquid radioactive waste.


G A S E O U S W A S T E S

Primary System Vents, Sampling System, Pressurizer- Relief Tank

LIQUID WASTES

H YD RO G ENA T ED DRA INS

Primary Coolant Letdown Sampling System Primary Component Drains

AERATED D R A IN S

Uncontaminated Drains Chem. Addition Tank Safety Injection System

Contaminated Drains Purif ication System Sample Sink Demineralizer Drains Containment Sumps

SOLID WASTES

C O M P R E S S IB L E WASTES

NON C O M P R E S S I B L E WASTES

Spent Resins

Spent Filter Cartridges

WASTE GAS HEADER

1

1

j

11

L -

V

SPENT RESIN SHIPPING CONTAINER

DRUMS

SPENT FILTER

SHIPPING CASK

(R«VMbl* Coth)

OFFSITE DISPOSAL

(6 Cartridge*)

i

175 FOOT PRIMARY VENT STACK

CIRCULATING WATER

DISCHARGE (372,000 gpm)

FIG. 9. Waste flowsheet at Connecticut Yankee Nuclear Power Station.


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See under USSR

UNION OF SOVIET SOCIALISTREPUBLICS

Mezhdunarodnaya Kniga Smolenskaya-Sennaya 32*34 Moscow G*200

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