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r/esno.56PROCEDURES AND DATA

Safety Assessmentfor the Underground Disposalof Radioactive Wastes

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From Safety Series No. 46 onwards the various publications in the series aredivided into four categories, as follows:

(1) IAEA Safety Standards. Publications in this category comprise the Agency’s safety standards as defined in “The Agency’s Safety Standards and Measures” , approved by the Agency’s Board of Governors on 25 February 1976 and set forth in IAEA document INFCIRC/18/Rev. 1. They are issued under the authority of the Board of Governors, and are mandatory for the Agency’s own operations and for Agency-assisted operations. Such standards comprise the Agency’s basic safety standards, the Agency’s specialized regulations and the Agency’s codes of practice. The covers are distinguished by the wide red band on the lower half.

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SAFETY ASSESSMENT FOR THE U N D ERG RO UND DISPOSAL

OF RADIOACTIVE WASTES


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

AFGHANISTANALBANIAALGERIAARGENTINAAUSTRALIAAUSTRIABANGLADESHBELGIUMBOLIVIABRAZILBULGARIABURMABYELORUSSIAN SOVIET

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Printed by the IAEA in Austria November 1981


SAFETY SERIES No. 56

SAFETY ASSESSMENT FOR THE UNDERGROUND DISPOSAL

OF RADIOACTIVE WASTES

INTERNATIONAL ATOMIC ENERGY AGENCY VIENNA, 1981


SAFETY ASSESSMENT FOR THE UNDERGROUND DISPOSAL OF RADIOACTIVE WASTES

IAEA, VIENNA, 1981 STI/PUB/590

ISBN 9 2 -0 -6 2 3 1 8 1 -2


FOREW ORD

This document is addressed to authorities and specialists responsible for or involved in planning, performing and reviewing safety assessments of underground radioactive waste repositories. It introduces and discusses in a general manner approaches and areas to be considered in making such safety assessments; its emphasis is on repositories for long-lived radioactive wastes in deep geological formations. It is hoped that this document will contribute to providing a base for a common understanding among the authorities and specialists concerned with the numerous studies involving a variety of scientific disciplines. While providing guidance, the document is also intended to stimulate further international discussion on this subject. It is the intention of the IAEA to develop more specific reports providing examples for the application of safety analyses for underground waste disposal.

The IAEA lias been active in the field of radioactive waste management for many years. In 1977, a draft proposal was prepared for a future IAEA programme on the underground disposal of radioactive wastes. An Advisory Group meeting, held from 30 January to 3 February 1978, confirmed this proposal and recommended that a set of guidelines be published concerning underground disposal of radioactive wastes. These guidelines are intended to cover the needs and interests of both developed and developing countries and to include the following subjects:

Generic and regulatory activities and safety assessmentsInvestigation and selection of repository sitesWaste acceptance criteriaDesign and construction of repositoriesOperation, shutdown and surveillance of repositories.

The present publication is part of this IAEA programme to develop guidelines on these subjects. A working draft was prepared with the assistance of a group of consultants who met in Vienna from 11 to 15 December 1978 and from 18 to 22 June 1979. The draft was revised by an Advisory Group meeting in Petten, Netherlands, from 8 to 12 October 1979 and examined by the Technical Review Committee on Underground Disposal of Radioactive Waste convened in Vienna from 10 to 14 November 1980.

A number of relevant publications have been or are being prepared within the IAEA’s programme on the safe underground disposal of radioactive wastes, addressing possible options for the disposal of high-, intermediate- and low-level radioactive wastes in deep, continental geological formations, in rock cavities at


various depths and in shallow ground. The following other publications have been recently prepared by the IAEA:

Site Selection Factors for Repositories of Solid, High-Level and Alpha-Bearing Wastes in Geological Formations, IAEA Technical Reports Series No. 177 (1977)

Development of Regulatory Procedures for the Disposal of Solid Radioactive Waste in Deep, Continental Formations, IAEA Safety Series No. 51 (1980)

Underground Disposal of Radioactive Wastes: Basic Guidance, IAEA SafetySeries No. 54 (1981)

Shallow Ground Disposal of Radioactive Wastes: A Guidebook, IAEA Safety Series No. 53 (1981)

Site Investigations for Repositories for Solid Radioactive Wastes in Shallow Ground, IAEA Technical Reports Series (to be published)

Site Investigations for Repositories for Solid Radioactive Wastes in Deep, Continental Geological Formations, IAEA Technical Reports Series (to be published),

Other appropriate IAEA publications prepared under the Radiological Safety Standards programme and the Nuclear Safety Standards (NUSS) programme might be consulted under the various related topics. In relation to the present document, the most important are:

Basic Safety Standards for Radiation Protection, IAEA Safety Series No. 9 (1967) (under review)

Principles for Establishing Limits for the Release of Radioactive Materials into the Environment, IAEA Safety Series No. 45 (1978)

Governmental Organization for the Regulation of Nuclear Power Plants: A Code of Practice, IAEA Safety Series No. 50-C-G (1978).

A list of other related background papers is provided in the Bibliography.

The IAEA gratefully acknowledges that the preparation of this publication was partially funded by the United Nations Environment Programme (UNEP) under its Project No. 0102-74-002 with the IAEA.


CONTENTS

1. INTRODUCTION .......................................................................................... 1

2. OBJECTIVES AND PROCEDURES OF SAFETYASSESSMENT.................................................................................................. 22.1. Objective of underground disposal ..................................................... 22.2. Safety and radiation protection requirements .................................. 32.3. Objective of safety assessments of underground disposal................ 42.4. Phases in safety assessments of underground disposal .................... 4

3. DISPOSAL SYSTEM DESCRIPTION ......................................................... 6

3.1. Disposal concepts ................................................................................ 73.2. Waste types .............................................................................................. 83.3. Host rock types ..................................................................................... 8

4. SAFETY ASSESSMENT APPROACH.......................................................... 9

4.1. General descriptions ............................................................................ 94.2. Modelling requirements ....................................................................... 114.3. Data requirements ................................................................................ 124.4. Model validation ..................................................................................... 124.5. Methods application ........................................................................... 13

4.5.1. Concept evaluation .................................................................. 134.5.2. Site and design evaluation ..................................................... 134.5.3. Licensing ..................................................................................... 13

5. SCENARIO ANALYSIS ................................................................................. 145.1. Identification of phenomena .............................................................. 145.2. Deterministic analysis ........................................................................... 155.3. Probabilistic analysis ........................................................................... 17

5.3.1. Fault/event tree analysis ......................................................... 175.3.2. Monte Carlo/Markov chain analysis ....................................... 18

6. CONSEQUENCE ANALYSIS ........................................................................ 19

6.1. Outline of procedure ........................................................................... 196.2. Areas of consequence modelling ......................................................... 21


6.2.1. Inventory................................................................................... 216.2.2. Release from the waste form ............................................... 226.2.3. Release from the repository................................................... 226.2.4. Transport within the geosphere .......................................... 226.2.5. Transport through the biosphere to man ............................ 236.2.6. Evaluation of doses ................................................................. 24

7. SAFETY ASSESSMENT................................................................................ 24

7.1. Combination of scenario and consequence analysis results .......... 257.2. Sensitivity and uncertainty analysis.................................................... 26

7.2.1. Objectives ............................................................................... 267.2.2. Method description ................................................................. 26

7.3. Evaluation and application of results ............................................... 277.4. Documentation of results ................................................................. 28

Glossary ..................................................................................................................... 31

Bibliography ............................................................................................................ 37

Drafting and reviewing bodies 39


1. IN TRO DUCTIO N

Nuclear energy is playing a continuing and increasing role in electricity generation in many countries. Therefore, these countries are faced with adopting appropriate systems for the management and disposal of the radioactive wastes. These wastes vary greatly in physical/chemical form and isotopic content. While various concepts have been proposed and are being studied, it is generally agreed that disposal underground, with the wastes appropriately immobilized, is an adequate way of providing the necessary protection for humans and the environment.

Underground disposal means the emplacement of radioactive waste materials in the terrestrial subsurface without the intention of retrieval. Accordingly, underground disposal concepts can range from shallow ground burial to emplacement in very deep bore-holes. Most effort is currently focused upon mined repositories in deep (e.g. up to 1000 metres), continental geological formations, and this emphasis is reflected in the present document.

Safety assessments are necessary to estimate the expected performance of a system considered for underground disposal of radioactive wastes and to compare it with acceptability criteria, both for the operational and post-operational phases; they are useful also to identify the potential significance of possible improvements of the system. They are important in every phase in system development: system selection; site confirmation; repository design, construction, operation, shutdown and sealing; and licensing processes relevant to these phases. Thus, safety assessments are a tool to predict the probable consequences of creating a waste repository, to compare the consequences with acceptability criteria (as defined in the Glossary), and to present the results for judgement by the appropriate bodies.

Safety assessments proceed from generic to site-specific studies. Generic assessments can be useful for making programmatic decisions regarding the choice of a disposal concept and the appropriate use of available resources. Generic assessments may also be helpful in gaining recognition of the feasibility of a disposal concept. Site-specific assessments are necessary for decisions affecting siting, design, and licensing for construction, operation, shutdown and sealing of a repository.

Safety assessments are performed by:

(a) working backward from acceptability criteria to determine performance specifications on system components (e.g. thickness of emplacement geology, emplacement spacing, canister corrosion rate, etc.);

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(b) working forward from the characteristics of systems to determine and compare their expected performance with acceptability criteria.

These two approaches are complementary and used in an iterative process. During this process the effects of potential improvements of the overall system of natural barriers and man-made provisions can also be evaluated relative to an ‘as low as reasonably achievable’ solution for the future consequences of the underground waste disposal, taking present-day economic and social factors into account.

Performance of such safety assessments requires an interdisciplinary team of specialists to cover the many technical aspects, working throughout all phases of repository planning, designing and licensing activities.

The purposes of this document are to introduce:

(a) overall safety assessment approaches that can be useful during all phases of disposal system development, including both generic and site-specific studies;

(b) general methods that can be applied within that overall approach.

Subsequent documents to be published by the IAEA will describe specific safety assessments and present further details on the methodologies used and the status of this rapidly developing technology.

2. OBJECTIVES A N D PROCEDURES OF SAFETY A SSE SSM E N T

2.1. Objective of underground disposal

The basic objective of underground disposal of radioactive wastes is to isolate them from man’s environment to ensure that any subsequent return of radionuclides to the environment will not result in undue radiation exposures to man. The necessary degree of isolation depends on the actual radionuclide content of the particular waste under consideration. The disposal system, including the waste forms, containers, engineered barriers, repository and the surrounding geological formation, etc., must be selected and designed to provide this appropriate isolation (see Section 3).

Until acceptability criteria for underground waste disposal systems have been defined by the national regulatory body and/or international bodies, an appropriate isolation may only be defined by an iterative and comparative assessment process in which the future consequences of a specific underground waste disposal facility will have been minimized according to an ‘as low as reasonably achievable’ principle.

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The basic radiation protection recommendations in ICRP Publication 26 and the IAEA Radiological Safety Standards1 apply to underground disposal of radioactive wastes. In some countries work is under way on the preparation of acceptability criteria and regulations2 for underground disposal. Some of the major topics under consideration within the process of criteria development are:

The degree of isolation, which is waste-type and time dependent

The overall radiological impact assessment

The extent of optimization.

The basic ICRP recommendations can be stated as follows:

(a) Justification — No practice resulting in human exposure to radiation should be authorized unless its introduction produces a positive net benefit. (The production of radioactive wastes is an unavoidable consequence of the generation of nuclear power and can therefore only be justified within the scope of an overall cost-benefit justification of the complete nuclear fuel cycle.)

(b) Optimization — All exposures should be kept ‘as low as reasonably achievable’, economic and social factors being taken into account. This requirement implies that the detriment from a practice should be reduced by protective measures to a value such that further reductions are not worthwhile in comparison with the efforts required to accomplish them.

(c) Individual dose limitation — The dose equivalent to individuals fromall practices (except those specifically excluded) should be less than the appropriate dose limits.

In the application of these recommendations, it must be recognized that many present-day practices might give rise to dose equivalents that will be received by future generations. This factor should be taken into account so that present and likely future practices would not be liable to result in radiation exposures to future generations higher than those acceptable to present generations.

2 .2 . S a fe ty a n d ra d ia tio n p ro te c tio n re q u ire m e n ts

1 “ Basic Safety Standards for Radiation P ro tection” , IAEA Safety Series N o.9 (1967) (under review).

“ Principles of Establishing Limits for the Release of Radioactive Materials in to the E nvironm ent” , IAEA Safety Series No.45 (1978).

2 See also “ Developm ent o f Regulatory Procedures for the Disposal o f Solid Radioactive Waste in Deep, C ontinental Form ations” , IAEA Safety Series N o.51 (1980).

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The ICRP individual dose limitation, like the other recommendations, applies only to planned or controllable situations which are expected to occur. For underground disposal these situations will vary with the disposal concept and the choice of site and host rock. For situations resulting from disruptive phenomena, the ICRP recommendations are not applicable and acceptability criteria need to be defined. The establishment of these acceptability criteria is a matter for the appropriate national regulatory body.

During the operational phase, occupational and public radiation exposures must also be kept in accordance with the ICRP recommendations and the regulations the national regulatory body may set. In addition to the radiation protection requirements, any other regulations and criteria regarding the conventional safety and non-radiological environmental protection aspects of a waste repository must be met.

2.3. Objective of safety assessments of underground disposal

It is the objective of safety assessments to analyse the expected performance of safety-related features of the underground disposal system, in particular the possibilities of any return of radionuclides from the repository to man; to compare the results of these safety analyses with acceptability criteria; and to make relevant judgements regarding the acceptability of the system and planned activities.

For a given repository a set of safety assessments will be likely to include generic studies as well as site-specific studies. The results of these assessments may be used by the appropriate regulatory body for the following purposes:

(a) To determine whether a site is suitable and a repository can be constructed, operated, shut down and sealed in compliance with relevant acceptability criteria

(b) To evaluate whether associated protection/prevention systems and measures (containment, barriers, stability of system, etc.) will function as assumed

(c) To determine necessary ‘authorized limits’ and specifications for design, operation and surveillance of the repository (waste conditioning requirements, temperature limits, radioactivity levels, etc.)

(d) To verify compliance with conventional safety criteria and criteria for environmental and resource protection

(e) To evaluate the overall radiological ‘impact’ due to the disposal.

2.4. Phases in safety assessments of underground disposal

Safety assessments are required for all the phases in the life of a waste repository. Figure 1 illustrates how safety assessments may interact with the other studies and activities necessary in the development of a waste disposal

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FIG. I. Safe ty assessments in waste disposal con cep t d ev e lo p m e n t and site se lection (taken f ro m "Underground Disposal o f R adioactive Wastes: Basic Guidance", Sa fe ty Series No. 54, IAEA).

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repository system. Generic assessments are required, in particular, during the general scientific and engineering evaluations and surveys to define probable concepts, host rocks, and suitable areas. They proceed then through preliminary to detailed site-specific assessments for preliminary site selection and the confirmation of the site and disposal concept and design. Subsequent assessments are required for the construction, operational and post-operational phases.

The operational phase covers the time during which wastes are being accepted for emplacement and during shutdown and sealing of a repository. During this time occupational exposures with possible public exposures occur from direct irradiation during handling and transportation and possibly from effluent releases. Attention should also be devoted to potential accidents following equipment failures or human errors. Thus, safety assessment for the operational phase of a repository is similar to that for operation of other types of nuclear installations.

Consideration must also be given to conventional safety requirements during the construction and operational phases. It has to be demonstrated that the construction and operation of a repository can be performed in accordance with the relevant safety rules (e.g. regulations regarding the safety of the mining activities); however, these aspects are not discussed further herein.

The present document emphasizes the special characteristics of radiological safety assessments covering the post-sealing phase; these safety assessment are made during the siting, design and licensing phases for a repository. For the postsealing phase the main objective is the prediction of radiation doses to the general public from natural, human, and waste- and repository-caused processes, with considerations extending far in the future.

3. D ISPO SA L SYSTEM D ESCRIPTIO N

For safety assessment purposes the overall system to be analysed can be described as a combination of the following components:

Waste type, waste form and, where applicable, stabilizer, container, overpack and/or migration retardantRepository and its engineered barriers (including backfill of excavations, bore-hole seals, and shaft seals)Geosphere (host rock and surrounding material and, where applicable, interstitial fluids in the host rock, deep groundwater and natural resources) Biosphere (soil, surface waters, shallow aquifers, atmosphere and biota).

The elements in the first three components represent barriers. The role of these barriers is to prevent or delay the initiation of radionuclide release from the waste,

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to distribute the release over time and retard transport of radionuclides through the geosphere and into the biosphere or the human environment.

Depending on the type of host rock available, the type of underground disposal concept chosen, and the type and form of waste to be disposed, the overall repository system will vary; not all of the components mentioned above will be relevant or required. The repository system designers therefore have considerable flexibility in choosing the components and thus optimizing their overall system design.

More detailed descriptions and considerations of these factors can be found in the IAEA document “Underground Disposal of Radioactive Wastes: Basic Guidance” , IAEA Safety Series No. 54 (1981).

3.1. Disposal concepts

Appropriately conditioned wastes may be disposed of in various types of underground repositories, dependent upon the types of wastes. The generic options are:

Emplacement of solid wastes at shallow depth, with or without engineered barriers, above or below ground surface, where the final protective covering is of the order of a few metres thick

Emplacement of solid wastes in mined repositories especially designed and excavated for waste disposal at suitable depths in continental geological formations, such as salt, crystalline rocks, or argillaceous formations

Emplacement of solid wastes in man-made or natural rock cavities at various depths

Injection of self-solidifying fluids containing wastes into induced fractures within low-permeability strata, including hydraulic fracturing

Injection of liquid or gaseous wastes into isolated porous and permeable strata.

Shallow ground disposal is described in the IAEA document, “ Shallow Ground Disposal of Radioactive Wastes: A Guidebook” , IAEA Safety Series No.53 (1981). It differs considerably from the other two concepts for underground disposal of solid wastes. Because emplacement is in the region of the biosphere, the containment and isolation of wastes in shallow ground repositories require a period of administrative control until the radionuclides decay to acceptable levels. Thus, although the objectives of safety assessment for shallow ground disposal do not differ from those for deep geological disposal, the methods used can be quite different for the following reasons:

The safety assessment covers a much shorter period of time because of the types of wastes disposed of in such facilities

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The containment barrier has greatly reduced dimensions

Breaching of the containment caused by natural or man-caused events ismore probable (flooding, land movement, erosion, excavation, etc.)

Remedial actions are conceivable.

In this report emphasis is placed on safety assessment methods for mined repositories in deep geological formations. However, the application of these methods to shallow ground disposal should be possible provided that the above- mentioned differences are taken into account.

Disposal in man-made or natural rock cavities is physically more similar to disposal in deep mined repositories in geological formations than shallow ground repositories. However, because of the proximity of some cavities of the earth’s surface and the potential for intrusion, safety considerations for this type of repository also resemble those for shallow ground repositories, listed above. Thus safety assessments for rock cavities will include features used for both the other concepts for underground disposal of solid wastes.

Deep bore-hole disposal, such as the last two generic options listed above, is currently used in some countries for the disposal of certain types of liquid waste. Safety assessment for this type of underground disposal may similarly make use of the methods described, provided that appropriate modifications are made to suit the specific requirements of these disposal concepts.

3.2. Waste types

The waste types under consideration for underground disposal can vary in form, in radionuclide content and composition, and in conditioning prior to disposal. Descriptions can be found in the IAEA documents mentioned earlier.

Different types of wastes require different treatment and conditioning prior to disposal in order to reduce volumes, to immobilize and contain the radionuclides and to increase safety and ease of handling during disposal operations. Safety assessments may lead to choices of different conditioning requirements and disposal techniques for different types of waste.

3.3. Host rock types

The host rock plays a very important role in the overall repository system. Different geologic formations are considered as potential host rocks for repositories. They include:

Evaporites (e.g. rock salt)Other sedimentary rock deposits (e.g. clay)Igneous and metamorphic crystalline rocks (e.g. granite).


Descriptions of the host rock types under consideration can be found in the IAEA Technical Reports Series No. 177, “ Site Selection Factors for Repositories of Solid High-Level and Alpha-Bearing Wastes in Geological Formations” , (1977).

The host rocks differ greatly in structure and properties and therefore exhibit different advantages and disadvantages with respect to the design of a repository system. Engineered barriers can be provided where necessary to cope with certain disadvantages of the host rock in order to make the overall system acceptable and/or to optimize that system.

4. SAFETY A SSE SSM E N T APPROACH

4.1. General descriptions

The process of safety assessment for underground disposal requires analysis of the combined effects of phenomena which could reduce the isolation effectiveness of a repository. As a first step, a concept must be developed for the total system, its components and interrelationships. This involves a preliminary characterization of the conditioned waste, the proposed repository design, the host system and the properties and phenomena having a potential influence on the overall waste disposal system and/or on specific components of this system.

Various classification schemes have been used or proposed for subdivision of a safety assessment into components. The classification scheme used in this report differentiates the following two major components:

Scenario analysis involves identification and definition of phenomena which could initiate and/or influence the release and transport of radionuclides from the source to man. Thus, scenario analysis provides initial and boundary conditions for subsequent consequence analysis.

Consequence analysis involves estimation of the subsequent transport of radionuclides from the source to man and the resulting radiation doses, using the system descriptions derived from the scenario analysis.

Comparison of the results of the biosphere consequence analysis with relevant acceptability criteria completes the safety assessment. Figure 2 illustrates these components of a safety assessment and their interactions. Iterations within specific analysis steps or complete iterations of the complete safety assessment process are normally performed.

Other classification schemes are possible for the categories of analysis. For example, it would be possible to use a spatial division as follows:

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FIG.2. Sa fe ty assessment com pon en ts and their interactions.


R epository analysis: this includes all the modelling and data collection required for the processes occurring within the excavated volumes. It includes, for example, container corrosion, waste leaching, stability of excavations, groundwater access to the repository, etc.

Geosphere analysis: this includes, for example, the thermal loading, the assessment of the geohydrological regime, and geochemistry studies.

Biosphere analysis: this includes consideration of all the processes which are relevant in the surface and near-surface environment.

In addition, there would be a division amongst disciplines which overlays with the spatial division, for example, as follows:

H ydrogeology: the analysis of water movement through all the spatial divisions.

Chem istry: the analysis of the geochemistry and chemistry in each of the spatial divisions.

Therm al and m echanical analysis: the analysis of heat transfer, stress distributions and effects on fracture populations.

The general approach with most classification schemes is to analyse the total repository system as it is expected to exist and to analyse the effects of variability in its defining parameters. Superimposed upon these analyses are the analyses of the probability and consequences of major disruptive events.

4.2. Modelling requirements

Safety analyses require the development of models which quantify the ways signficant phenomena occur. A model is a mathematical representation of a real system which is sufficiently simplified and compact to be amenable to quantitative analysis. Models can be classified into deterministic and probabilistic models. Individual or sub-models can be combined to give system models. In some cases, deterministic and probabilistic models may be combined in the same system model.

Scenario analysis models are used for analysing potential phenomena which might change the state of the system. For example, faulting through or in proximity of a repository might change the permeability of the formation and the pattern of groundwater flow; or with respect to glaciation it would be necessary to estimate the effect of the overburden pressure on the degree of fracturing of the geologic formation and the flow of groundwater.

Consequence analysis models are directed towards the estimation of the rate of release and transport of radionuclides within the system defined by the scenario analyses and the corresponding radiation doses to man. These models analyse

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hydrogeological and hydrologic systems, including chemical effects, thermal effects and transport and uptake of radionuclides. Examples include models for leaching from wastes, for movement of radionuclides through groundwater systems and for transport through the biosphere to man.

4.3. Data requirements

For safety analyses of particular systems the following data are tj/pically required:

Waste characterization: (composition versus time, quantity, heat generationversus time, etc.)

C ontainer characteristics: (mechanical, chemical, etc.)

R eposito ry characteristics: (dimensions, backfill/buffer, structuralmaterial, etc.)

Geosphere characteristics: (geology, hydrogeology, geochemistry, etc.)

Biosphere characteristics: (atmosphere, aquatic, terrestrial, demographic, etc.)

Data can be collected for components of the total system from existing data files and literature, laboratory experiments, in-situ tests, and field observations; the utility of the data should be considered in the context of potential human activities, also. An important and unique aspect is that the data need to be applicable for times far in the future. In some cases long-term processes may be accelerated in experiments that provide data which can be applicable to long-term conditions.In other cases conceivable variations in data (e.g. due to climatic change) can be accounted for by variation analyses.

4.4. Model validation

The value of models used in safety analyses is a function of the degree to which: (a) the processes treated in the model represent the actual interactions in the real system; (b) the data used in the model represent good estimates of the actual data which define the real system. The central problem of safety assessment for the post-sealing phase of a waste repository is that demonstration is only possible in the short term whereas the problem to be solved has a longterm component.

Clearly, predictions of the evolution, over thousands of years, of man and the environment must be based on assumptions. Prediction into the distant future is a special difficulty of the analyses; on the other hand, the interacting processes themselves are not of unprecedented complexity. Just as many interactive processes are considered in, for example, simulation of the behaviour of

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chemical plants. However, in the waste repository system, the sequence of processes is slower. A further difference is in the area of validation. For a chemical plant data can be collected on the behaviour of the complete system and so the whole analysis can be checked, although difficulties can still arise in, for example, verifying predictions of low failure rates.

For a repository system no such access to overall system behaviour is available. However, data can be collected for individual components of the system and so models can be validated for these components. For example, radionuclide transport models can be checked against laboratory measurements, in-situ tracer tests or observations of the movement of existing radionuclides in the environment. A limited number of accelerated tests can be carried out which help resolve the problem of the long time-scale by causing the process to occur much more rapidly. However, for example, if experiments are carried out at increased temperature to hasten the rate of chemical reaction, much caution is needed in extrapolating the results of such experiments to the actual conditions in a repository, where such high temperatures may never be reached.

A final stage of validation to be justified and verified is the linking together of individual concepts and models into a comprehensive safety assessment of a system.

4.5. Methods application

4.5.1. C oncept evaluation

Concept evaluation involves the generic assessment of the acceptability and effectiveness of a proposed repository system. No consideration is given to site selection, but the value of a concept evaluation is increased considerably if real ... data are used as much as possible. The results of a concept evaluation will enable the appropriate bodies to judge: (a) whether the developed methodology is sufficient, and (b) whether a proposed repository is likely to be acceptable.A decision can then be made on proceeding to the evaluation of actual sites.

4.5.2. S ite and design evaluation

Site and repository design evaluation involves a safety assessment consideration with respect to the acceptability of a proposed waste repository at a proposed site. It provides a basis for selecting from different designs and sites.

4.5.3. Licensing

Most countries are choosing to regulate their disposal programmes through licensing actions by a regulatory body whose purpose it is to review, certify and

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ensure the safety of the selected repository system. This requires safety assessment at several stages in the licensing and implementation process. Detailed guidance concerning the licensing process may be found in “Development of Regulatory Procedures for the Disposal of Solid Waste in Deep Continental Formations” ,IAEA Safety Series No.51 (1980).

5. SCENARIO ANALYSIS

As indicated above, scenario analysis involves the identification and definition of phenomena, and their interactions, which could initiate and/or influence the release and transport of radionuclides from the source to man. Following a discussion of the identification of phenomena, the techniques for defining the phenomena explicitly and quantitatively are discussed: deterministic analysis and probabilistic analysis. Although these techniques are discussed separately, the distinction between them is rather artificial since probabilistic analysis inevitably involves deterministic analysis and deterministic analysis also has some probabilistic components (e.g. statistical distributions of parameter values may need to be considered). Both techniques of analysis are generally used in a complementary way in comprehensive safety assessments to model simultaneously discrete events and continuous physical processes.

5.1. Identification of phenomena

Several types of occurrences could lead to the release of radionuclides and in some cases enhance such releases. To analyse the release scenario, it is necessary to identify the phenomena which are relevant. These phenomena could be due to:

(a) Effects of natural processes and events (e.g. groundwater flow, erosion, faulting, etc.)

(b) Effects of human activities (e.g. alterations of hydrology, mining, drilling, etc.)(c) The combined effects of the waste and repository (e.g. thermal, chemical,

mechanical, radiological, etc.).

A schematic representation of an underground repository system and these important phenomena is presented in Fig.3.

In choosing phenomena relevant to a particular instance, it is useful to have a checklist. A suggested checklist of phenomena is presented in Table I. Although this checklist may not contain all phenomena relevant to waste repositories, and the organization is somewhat arbitrary, it provides a comprehensive view of potential phenomena.

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Hum an a c tiv itie s

- A lte ra tio n s o f b iosphere

- A lte ra tio n s o f h y d ro lo g y- M in ing and d r il l in g

N atura l processes

- G ro u n d w a te r f lo w

- E rosion

- F au lting /S e ism ism

- U p lif t- V o lcan ism

W aste and re p o s ito ry e ffects

- Therm al- Chemical- M echanical- R ad io log ica l

A tm osphere

__S ed im entary

layers

(h ig h ly variab le)

B iosphere

G eosphere

- N atura l resources

B a c k fille d tunne l

C o n d itio n e d a lpha-bearing wastes B a ck fille d s tab ilize r

D r ille d holes

C o n d itio n e d h igh-level wastes (Engineered barriers)

FIG.3. Schem atic diagram o f underground reposi to ry sy s tem (after sh u tdow n and sealing).

The challenge of the scenario analysis activity is to identify those phenomena which could make a significant contribution to the overall radiological impact and to eliminate those which would make a negligible contribution because their probabilities of occurrence or their potential consequences would be insignificant. For example, phenomena such as undetected past intrusion and changes in sea level may be rejected because they have been taken into account in selecting the site; the chosen depth of the repository may enable phenomena such as meteorite impact and stream erosion to be discounted.

5.2. Deterministic analysis

Deterministic analysis is the classical systems analysis technique for studying system behaviour using the laws of science and engineering. This technique requires that explicit relationships exist between all processes at work on and within a system. This kind of analysis has two parts, the determination of whether certain occurrences are possible and the prediction of the resulting conditions when those occurrences which are possible happen. The usual method of accomplishing the first part is stability analysis; the method for accomplishing the latter is perhaps best named science and engineering analysis. Stability analysis is a deterministic

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TABLE I. PHENOMENA POTENTIALLY RELEVANT TO RELEASE SCENARIOS FOR WASTE REPOSITORIES

Natural processes and eventsa

Climate change Hydrology change Sea level change D enudation Stream erosion Glacial erosion Flooding Sedim entation Diagenesis DiapirismFaulting/Seismicity Geochemical changes

Fluid interactions Groundw ater flow Dissolution Brine pockets

Human activi ties

Undetected past intrusion Undiscovered boreholes Mine shafts

Im proper design Shaft seal failure Exploration borehole seal failure

Im proper operationIm proper waste emplacement

Transport agent in troduction Irrigation ReservoirsIntentional artificial groundwater

recharge or withdrawal Chemical liquid-waste disposal

Waste and reposi tory effects

Thermal effectsDifferential elastic response Non-elastic responseFluid pressure, density, viscosity, changes Fluid migration

Chemical effects CorrosionWaste package — rock interactions Gas generation Geochemical alterations

Uplift/SubsidenceOrogenicEpeirogenicIsostatic

U ndetected features Faults, shear zones Breccia pipes Lava tubes Intrusive dykes Gas or brine pockets

Magmatic activity Intrusive Extrusive

Meteorite impact

Climate change (including climate control)

Large-scale alterations o f hydrology

Intentional intrusion WarSabotage Waste recovery

Inadvertent future intrusion Exploratory drilling Archeological exhum ation Resource mining (mineral, water,

hydrocarbon, geothermal, salt, etc.)

Mechanical effects Canister movement Local fracturing

Radiological effectsMaterial property changes RadiolysisDecay product gas generation Nuclear criticality

a Explanations of natural phenomena are provided in IAEA Technical Reports Series No. 177, “Site Selection Factors for Repositories of Solid High-Level and Alpha-Bearing Wastes in Geological Form ations” (1977).

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technique for analysing whether important properties of a system of interest remain unchanged if the system is disturbed. Unstable situations identified from this analysis are then subjected to science and engineering analysis to determine the effect of their occurrence on the system.

For the application of deterministic analysis techniques, the processes of the system must be expressed in quantitative mathematical form and data must exist to evaluate the physical constants and boundary conditions. This requires that the system of interest be well characterized. Deterministic analysis seems well suited to analysing the effect of processes on the repository system. However, the technique is poorly suited to analysing the events which are difficult to model by continuum mathematics. Furthermore, since deterministic analysis requires a well-characterized system, it has limitations for the natural geologic processes. Thus, deterministic analysis is not a comprehensive technique, but rather is a specialized method for analysing a portion of the scenario analysis problem.

Examples for application of deterministic analyses in scenario analysis include:

(a) The estimation of the life of a container given the incident water composition and flow rate, the container properties and relevant experimental corrosion data

(b) The estimation of effects of glacial loading and rebound on fracture distributions in hard rock by the use of stress analysis and linear elastic fracture mechanics techniques.

5.3. Probabilistic analysis

Probabilistic analysis involves a set of statistical techniques for studying effects. Parameters whose values are uncertain, events whose occurrences are random, and features which may or may not be present can be treated statistically. For example, a distribution function of permeability versus probability may be constructed for a host rock formation; the probability or frequency of occurrence of a meteorite impact may be estimated; and the probability of the existence of a fracture undetected by drilling can be calculated.

Various methods are then available for considering how the probabilistic variations in components of a system act together to cause variation in the system as a whole. These include fault/event tree analysis and Monte Carlo/Markov chain analysis.

5.3.1. Fault /even t tree analysis

The fault tree and event tree analysis techniques are the conventional methods for system reliability and probabilistic safety analysis. In this approach, system failure logic is graphically displayed in treelike structures. Computer-aided methods

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are available for analysing these structures to determine both qualitative and quantitative aspects of system reliability and probabilistic safety performance.The precise application of fault/event tree analyses requires thorough knowledge of the system under study and explicit interrelationships among the various components which comprise the system.

Fault tree analysis is a deductive probabilistic technique by which the component failures leading to system failure can be logically deduced. Application of the technique yields combinations of basic events whose occurrence causes the undesired system failure events. Fault tree analysis starts with careful definition of the failure events and systematically diagrams backward to identify the events or combinations of events that could cause the failure event to occur. The logic is displayed graphically using Boolean logic in a treelike structure. The process stops when the analysts define events that are either not amenable to further resolution or where it appears there is no need for additional definition.

Event tree analysis is an inductive probabilistic technique which reverses the fault tree approach by starting with the basic initiating events and working forward in time in order to display their logical propagation to system failure events. Event trees diagramatically illustrate the alternative outcomes or consequences of specified initiating events. The trees provide a pictorial representation of outcomes and provide the basis for a quantitative risk assessment of the various sequences which arise from the initiating event.

Because of its binary logic, the fault/event tree approach is well suited for analysing events which can either occur or not. For analysis of slow continuous processes, this probabilistic approach seems less suited. However, it can be utilized in this case by dividing the processes into classes characterized by different rates, each one being distinguished on the basis of a particular probability value.

5.3.2. M on te Carlo/M arkov chain analysis

Monte Carlo and Markov chain techniques are stochastic methods of simulation analysis. These methods are generally used for systems that cannot be described in a deterministic manner either because the system is too complex to be described in abstract mathematical form or because the mathematical equations cannot be solved analytically.

Monte Carlo analysis (as applied in scenario analysis) begins by listing the basic system events and estimating their probabilities of occurrence. The analysis then steps through time, assuming occurrence of these events according to their estimated probability distributions until a system failure occurs. After the simulation is performed a large number of times, a distribution of failures can be determined.

Markov chain analysis begins by specifying the possible states of the system and determining the probabilities for transition between those states. Implicit

18


in the analysis is the assumption that the transitional probability between two states is independent of earlier states of the system. The analysis then steps through time, assuming probabilistic occurrence of the transitional events until a system failure occurs. Rather than simulating the system repetitively as in Monte Carlo analysis, Markov chain analysis uses extensive mathematical formalism to reduce the Markovian processes to analytical functional forms.

Both techniques are well suited to the quantitative analysis of discrete events, and because of their stochastic nature they are also well suited to analysing the complex interactions of these events. A disadvantage of Monte Carlo analysis is the large number of simulations expected to be necessary to obtain adequate failure Distributions for low-probability events and their combinations. Monte Carlo analysis can be generalized to consider the continuous processes present in geologic isolation systems by updating the state of the system for their effect at each time step. Markov chain analysis, on the other hand, does not appear applicable to the continuous processes because the states of the system must be specified a priori in order to apply the method. Furthermore the transitions between states of a geologic system seem likely to be dependent on the transitions between earlier states, in violation of the basic assumption of Markovian processes.

6. CONSEQUENCE ANALYSIS

The ultimate aim of consequence analysis is to determine the radiation doses to man resulting from the disposal of radioactive waste. Doses to man can occur during the operational phase and the post-operational phase of a repository. During the operational phase, occupational exposures and exposures of the general public will need to be assessed. The consequences of an accident during this phase must be considered and account taken of appropriate remedial action, including possible abandonment and premature sealing of the repository or a portion thereof.

In the post-sealing phase, control is relinquished over very long time-scales.It is therefore necessary to analyse consequences of the scenarios, developed previously, using mathematical models. The main features of the relevant models are described in sections 6.2 .1 to 6.2.6. The outline of the procedure is given in section 6 . 1 .

6 .1. Outline of procedure

The analysis of the radiological consequences involves calculations of the release, dispersion and transport of radionuclides from the waste form through engineered barriers, the repository, the geosphere, and the biosphere and finally

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I-----------------------------------------------------------J

] MODEL FOR TRANSPORT AND/OR PHYSICAL/(BIO-)CHEMICAL PROCESSES

CONNECTION BETWEEN TWO M O DELS (INPUT-OUTPUT RELATION)

FIG.4. Rela tionships in consequence modell ing.

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calculations of radiation doses to man. These calculations lead to the estimation of individual and collective doses.

The consequence analysis begins with analysing the release of radionuclides from the waste form and the repository. Temperature distributions, mechanical stress conditions, radiolytic effects, corrosion and sorption are the major factors involved. Subsequently, the geochemical and geohydrological processes in the host formation are analysed to estimate the rate of transfer of radionuclides through the geosphere to the biosphere. Finally, the processes of biological uptake by man and resultant exposure are analysed to arrive at the doses to man.

Making the necessary calculations requires that the observed systems be mathematically described by models. It may be possible to work with specific models for each process in the pathway from the waste to man, but integrated models are also useful. The models must, however, be linked together in such a way that the calculations can be performed reliably. Modelling may be for generic purposes or for site-specific purposes based on extensive field studies.The combinations of models used will depend upon the situation.

An idealized system of models consists of the following:Inventory of radionuclides in the waste Release from the waste form Release from the repository Transport through the geosphere Transport through the biosphere Uptake and doses to man.

However, not all components are required in every analysis. For example, human intervention may lead to direct exposure to the waste; i.e. branching may occur at the point of release from the repository. Figure 4 shows the relationships involved with detailed consequence modelling. In certain circumstances global models may need to be used.

6.2. Areas of consequence modelling

Although it is convenient to use a set of different models instead of one global model, it must be realized that there will be interactions between various models, and modelling requirements will depend upon the situation. There will be overlaps between models; for example, the use of deep groundwater as a source of fresh water will result in pathways in both the biosphere and geosphere models. Consistent assumptions must apply through all models.

6.2.1. Inven to ry

The contents of all significant radionuclides in the waste and the decay processes define the inventory for further calculations. The decay processes

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also imply ingrowth of daughter nuclides, some of which will be radioactive and have different chemical and physical properties. Consequently, the inventory changes with time.

The heat produced by radioactive waste, especially the high-level waste, and the resulting temperatures in the waste form and the near surroundings must be calculated and used as inputs for subsequent calculations.

6.2.2. Release fro m the waste fo rm

Apart from a few potential disruptive events the release of radionuclides from the waste can only occur as a result of leaching by groundwater. Human intervention would be an example of a disruptive event that might release radionuclides otherwise. The potential release processes may be quite different when disposal in salt, hard rock and clay are concerned. In hard rock and clay there is always some groundwater flow, although a formation can be chosen which has a very low permeability for water. In a salt formation the normal condition is one of an absence of water, although there is a possibility of water inclusions or a risk of water intrusion.

The waste form is protected from leaching until the containers fail. When water comes into contact with the waste form, leaching will start; the release is generally described by a leach-rate model. The rate of release is controlled by many parameters such as the composition of waste form, surface area in contact with water, temperature, water chemistry and water flow.

6.2.3. Release fro m the repository

Depending upon the design of the canister and waste form, it is possible that nuclides will not be released for a long time. When release does occur, other engineered barriers may be present in the repository to sorb and/or otherwise retard release of radionuclides from the repository. The influence of these barriers will need to be assessed. The significance of engineered barriers depends upon the situation under consideration.

For disposal of certain types of waste (e.g. unreprocessed spent fuel), attention must be given to the possibility of reconcentration of fissile materials leading to nuclear criticality.

6.2.4. Transport w ith in the geosphere

The main transport mechanism in the geosphere is migration with groundwater. The transport of the radionuclides in the geosphere is governed by the water flow and geochemical retardation effects. The water flow may take place in a more or less homogeneous material. Pronounced anisotropic conditions may

22


also prevail mainly in fractured rocks. For example, groundwater will generally follow the natural hydraulic gradient, but in fractured rocks it will also be controlled to some extent by local fracture patterns and fracture properties.

Transport in a permeable medium with water flowing through pores or cracks is normally described by uni-dimensional models for parametric studies and for detailed analysis of long migration paths. Multidimensional models are useful for selection of critical paths and for calculation of effects close to the repository. The type of calculations and the required precision should govern the model selection. When necessary, the models might also describe the flow- patterns in anisotropic and non-homogeneous systems.

For the transport of radionuclides with groundwater not only gradient flow but also diffusion, dispersion, thermal convection, turbulent flow and density flow may be taken into account. Most of the nuclides are retarded compared with the velocity of the water. Several physical and chemical processes may contribute to the retardation. The retardation factors vary among the radionuclides and are also dependent on surface properties, crack widths, water chemistry, etc. For site-specific situations some of these parameters may be determined by laboratory or field experiments.

Besides the above-mentioned groundwater transport modes, sometimes bulk transport of the waste together with parts of the surrounding formation (e.g. active diapirism of salt domes) has to be considered. In this case models based on the expected geological displacement are applicable. However, it is often very difficult to obtain enough geological information as input for this type of model. In the case of salt, bulk transport might also be induced by the temperature effects on the viscosity of the formation. Temperature and density effects on repository stability must also be considered where necessary.

6.2.5. Transport through the biosphere to man

Transport to the biosphere starts when radionuclides:

(a) Enter surface water(b) Enter drinking water wells(c) Come within the range of roots of plants, or(d) Arrive at the surface as a consequence of bulk transport.

Various pathways are taken into account for transport within the biosphere3.The relative importance of the pathways will depend upon the environment, the population habits and the radionuclides. The ingrowth of daughter radionuclides must be taken into account in biosphere modelling.

E x p o su re s to Critical Groups from Routine Releases: Generic Models and Param etersfor Assessing the Environmental Transfer of Radionuclides, IAEA Safety Series (to be published).

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Airborne radionuclides may irradiate the population in several ways. There may be external irradiation due to immersion in contaminated air or irradiation by material deposited on the earth surface. There may be also internal irradiation due to direct inhalation of contaminated air or ingestion of foodstuff or water contaminated by airborne material.

Terrestrial pathways lead to irradiation of man by three important routes: external irradiation from radionuclides transferred to the earth surface, internal irradiation from inhalation of resuspended activity, and ingestion of contaminated foodstuffs. Aquatic pathways by which man may be irradiated include drinking water from aquifers or rivers, ingestion of foodstuffs, and external irradiation from sediments.

For the long-lived radionuclides an extensive model describing the dispersion and transport on a regional or global scale is necessary. The globe may be divided into local, regional, intermediate and global zones. Each of these zones may be divided into compartments describing the transport and uptake in the major reservoirs.

6.2.6. Evaluation o f doses

External and internal irradiation are considered using the pathways described in section 6.2.5. Individual doses and collective dose commitments are calculated.4 The doses to be evaluated depend on the purpose for which the safety assessment results will be used. For some purposes it may be sufficient to evaluate maximum individual doses while for others detailed evaluations of collective doses and collective dose commitments may be required.

7. SAFETY A SSESSM E N T

The quantities usually obtained from the scenario and consequence analyses for a safety assessment include release rates and/or concentrations of radionuclides entering defined compartments of the system, the probabilities of occurrence of releases, and the potential doses to man. These doses can be compared with corresponding radiological safety standards and acceptability criteria. Frequently, the probabilities of the releases which cause the exposures are also evaluated along with the consequences to indicate the risks.

4 See Glossary for definition.

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With the approach presented in section 4, the results of the scenario analyses and consequence analyses can be combined by several simulation analysis methods, e.g. by traditional risk analysis, ‘worst-case’ analysis, etc.

In traditional risk analysis, a limited list of scenarios is produced by the scenario analysis studies. Both the probability or likelihood of occurrence and the corresponding doses to man are calculated for each scenario. These probabilities and consequences are multiplied together and the resulting risks are summed and integrated over time to yield a value for the total detriment. 5

An advantage of this method is the relatively small number of scenarios and consequences that must be analysed, and a disadvantage is the difficulty in considering distributions of the analysis parameter.

In ‘worst-case’ analysis, scenarios selected from the scenario analysis studies are those which represent the most severe release situations based on a number of pessimistic assumptions. The consequences of these ‘worst-case’ scenarios (for which occurrence the probabilities are assumed to be one) are calculated and presented as upper limits for possible future doses. If it can be agreed that the scenarios are indeed ‘worst-case’ and that the consequences are within acceptable limits, ‘worst-case’ analysis can be an efficient method of performing a safety assessment. However, if agreement cannot be reached on both conditions, a more detailed approach may be necessary. For example, agreement may be reached on a set of ‘conservative, but realistic’ scenarios that allows completion of meaningful safety assessments.

A major expansion of the set of ‘conservative, but realistic’ scenarios can also be useful for safety assessments. An extensive list of specific scenarios is produced by the scenario analysis studies. A given scenario may appear in the list many times but have different conditions of occurrence for each entry. If a sufficiently large sample of specific scenarios has been selected, each scenario will occur in the list in direct proportion to its likelihood or probability of occurrence. The consequences for all scenarios can be added together to obtain a value for the total detriment; however, this value will have no absolute meaning because no absolute probabilities will have been used in its determination and, therefore, it will be of limited use. An advantage of this approach is the ability to include distributions rather than single values for analysis parameters, but a disadvantage is the large number of scenarios and consequences that must be analysed. However, with this approach the performance of an uncertainty analysis is made easier.

7 .1 . C o m b in a tio n o f scen a rio an d co n seq u en ce analysis resu lts

5 A single product o f probability and consequence is not sufficient to compare a low- probability/high-consequence event with high-probability/low-consequence events because people are usually more concerned about high-consequence events.

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7 .2 . S ensitiv ity a n d u n c e r ta in ty analysis

7.2.1. Objectives

The quantities obtained by combining scenario and consequence analysis results are the expected values of individual radiation dose and the collective dose commitment. Of course, these quantities are complex functions of all the input data required for the modelling. Obtaining some of these data is difficult due to the complexity of the systems and the long time-period of interest for the safety assessment of underground repositories. It is, therefore, important to evaluate the overall waste repository system design and development using sensitivity and uncertainty analysis.

Sensitivity and uncertainty analysis can identify at any stage of the repository system design and development those scenarios and consequences, including the input data involved and mathematical models, which are relevant for the determination of the above-mentioned quantities and gives the basic information to enable:

(a) Estimation of the overall uncertainties and error bounds of the quantities(b) Judgement on which data and models further theoretical and experimental

effort should be focused.

Sensitivity and uncertainty analysis can be applied iteratively in conjunction with the results of continuing research activities to determine necessary adjustments and the directions for future research.

7.2.2. M ethod description

There are three approaches for sensitivity and uncertainty analysis:

Simple parameter variationStatistical uncertainty evaluationPerturbation analysis.

For simple parameter variation, the input parameters for the scenario and consequence analyses are varied arbitrarily about some standard or mean values within reasonable ranges. For each set chosen, a complete safety analysis is performed. For different sets of parameters this procedure is repeated. The results can be expressed in the form of a so-called sensitivity profile which expresses the percentage change of the corresponding dose per percentage change of the specific parameter under consideration. The sensitivity profiles may be combined with the error bounds on the input parameters to calculate an overall error for the corresponding dose, which is an estimate of the uncertainty.

26


For statistical uncertainty evaluation, various statistical techniques can be used to select parameters for the safety analysis. Eventually all available or estimated information on the input data has to be used, such as mean or probable values, variances, standard deviations, probability distributions or lower and upper bounds. A sufficiently large number of such selections is performed so as to result in an estimate of the overall uncertainty of the risk, such as mean values, variances, lower and upper bounds or at the maximum some probability distribution. Intermediate results may be sensitivity profiles, too.

Perturbation analysis is particularly applicable in the case of deterministic models described by differential/integral equations. It is based on the direct and adjoint solution of the problem- The only conditions are that the adjoint equation can be formulated and its solution can be reasonably interpreted. The adjoint solutions can be found, as for the direct equations, either by deterministic or stochastic methods. The perturbation analysis method is a well-known tool from the solution of differential/integral equations which may be appropriately applied to the set of equations used to model the processes important to the safety assessment of underground repositories.

7.3. Evaluation and application of results

The safety assessment approach in this document is applicable for:

Concept evaluationSite evaluationRepository design evaluationRepository system licensing.

The level of sophistication of the assessment technology and the quality of the assessment results increases as the repository system design and development proceed. The results of the safety assessments can be used by the appropriate bodies to make relevant comparisons and judgements. The acceptability of a repository system can be judged on the basis of ICRP recommendations and other international and national criteria and regulations which may apply.

Safety assessments include the calculation of occupational doses, individual doses to members of the public and collective doses to present and future generations. By use of established relations between doses and health effects (as given, for example, by the ICRP), health effects can be predicted. In the case of underground disposal of radioactive waste the long time-periods have to be considered when a judgement on the acceptability of a proposed repository system is made.

For purposes of optimization, best estimates of parameter data are important. When pessimistic or uncertain data have to be used, the definition of the degree of uncertainty is useful to aid in interpretation of the significance

27


of results. The results of sensitivity and uncertainty analyses therefore should be taken into account.

To put the safety assessment results for underground repositories into perspective for the generally interested public, appropriate comparisons can be made. Examples are:

(a) Comparison of estimated probabilities and consequences of radionuclide release scenarios with those for corresponding scenarios for other activities of the nuclear fuel cycle

(b) Comparison of estimated effects of nuclear power generation with those of other power generation alternatives

(c) Comparison of the long-term impact of nuclear wastes with hazardous chemicals in wastes from contemporary alternatives for power generation

(d) Comparison of estimated doses from repositories with doses for an equivalent period of time from naturally occurring radionuclides

(e) Comparison of calculated potential hazard against the hazard from naturally occurring ores.

Such comparisons will also be necessary for determining the acceptability of a given waste repository until specific criteria are set. Setting of acceptability criteria is a topic which is under intensive discussion. Thus, the design team and the safety analysts may be forced to keep abreast of this discussion and submit their documented results of an ‘as low as reasonably achievable’ approach in each of the successive steps of the overall safety assessment process for consideration by the regulatory bodies, who ultimately have the responsibility to develop the acceptability criteria.

In addition to the radiological issues, the nonradiological detriments for the underground repository system may be important considerations. For example, systems for underground repositories require significant construction work and mining operations. It is therefore important that the detriments from these aspects are also considered when different alternatives are being compared.

7.4. Documentation of results

Clear and effective documentation of safety assessment results is necessary for communication not only with the regulatory body but also within the implementing organization and with other interested parties. Such documentation is needed especially because safety assessments of waste repositories are not numerous or familiar documents and the technology is rapidly developing.

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Consideration should be given to including in a safety assessment report summaries with their contents oriented explicitly for the various readers.

The organization of a safety assessment report should allow for presentation of principally narrative information and summary data in the body of the report. Supporting detailed mathematical and numerical data should be supplied in appropriate separate appendices. The general organization of such a safety assessment report text would be quite similar to that of the present document.

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GLOSSARY

ALARA. “As low as reasonably achievable, economic and social factors being taken into account”. A basic principle of radiation protection taken from the Recommendations of the ICRP, ICRP Publication No. 26 (p. 3).

aquifer. A water-bearing stratum or formation below the surface of the earth that can furnish an appreciable supply of water for a well or spring.

barrier (natural or engineered). A feature which delays or prevents radionuclide migration from the waste and/or repository into its surroundings. An engineered barrier is a feature made by or altered by man; it may be part of the waste package or part of the repository.

biosphere. That portion of the earth’s environment inhabited by living organisms. It comprises parts of the atmosphere, the hydrosphere (ocean, inland waters and subterranean waters) and the lithosphere. (See also human environment.)

collective dose equivalent commitment (or collective effective dose equivalent commitment). (Effective) dose equivalent commitment multiplied by the number of individuals in the specified population. It is commonly expressed in units of man-sievert (man'Sv) or man-rem. (Note that the modifiers ‘incomplete’ and ‘truncated’ are sometimes used for specific purposes of radiation protection — the details of which must be clarly stated.)

conditioning of waste. Those operations that transform waste into a form suitable for transport and/or storage and/or disposal. The operations include converting the waste to a more stable form, enclosing waste in containers, and providing additional packaging.

consequence analysis, biosphere. A safety analysis that estimates potentialindividual and population radiation doses to humans, based on radionuclide releases and transport from a nuclear installation (e.g. a waste repository) to the human environment as defined by hypothetical release and transport scenarios.

containment. The retention of radioactive material in such a way that it is effectively prevented from becoming dispersed into the environment.

criteria. Standards on which a decision or judgement can be based. They may be qualitative or quantitative. Acceptability criteria are those criteria for a repository system acceptable to the regulatory body. Some Member States use terms such as ‘protection goals’ instead o f ‘acceptability criteria’.

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deterministic analysis. A classical technique for studying the behaviour of a system mathematically using the laws of science and engineering, provided that all system parameters, events and features are deterministically (as opposed to probabilistically) defined.

detriment. The mathematical expectation of harm to a population incurred from a collective dose equivalent commitment, taking into account not only the probabilities of each type of deleterious effect but the severity of the effect as well. Detriment, in general, also includes deleterious effects not associated with health, such as the need to restrict the use of some areas or products.

disposal. The emplacement of waste materials in a repository without the intention of retrieval.

disruptive event. An event (e.g. an earthquake, or meteorite impact) that disrupts a nuclear installation (e.g. a waste repository).

dose equivalent. The product of absorbed dose and quality factor and other modifying factors necessary to obtain an evaluation of the effects of irradiation received by exposed persons, so that the different characteristics of the exposure are taken into account. It is commonly expressed in units of sieverts (Sv) or rems.

dose equivalent commitment (or effective dose equivalent commitment). Forany specified decision, practice or operation, the infinite time integral of the per caput dose-equivalent rate for a specified population. The exposed population is not necessarily constant in numbers. It is commonly expressed in units of sieverts (Sv) or rems. (Note that this can apply over very long (geological) times and care must be taken to maintain perspective. The concept can be useful in making comparisons among alternatives but in a given case may have little meaning in an absolute sense).

event tree analysis. An inductive probabilistic technique that starts with hypothesizing the occurrence of basic initiating events and proceeds through their logical propagation to system failure events. The event tree is the diagrammatic illustration of the alternative consequences or outcomes of specified initiating events.

fault tree analysis. A deductive probabilistic technique that starts with hypothesizing and defining failure events and systematically deduces the events or combinations of events that could cause the failure events to occur. The fault tree is the diagrammatic illustration of the events in a tree-like structure.

generic analysis. A generalized analysis for a nuclear facility; for a waste repository it is for a type of host rock, as opposed to an analysis for a site-specific host rock.

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geosphere. That geological (solid) portion of the earth’s environment not normally inhabited by living organisms, synonymous with lithosphere.

host rock or host medium. The geological formation in which the repository is located.

human environment. Those portions of the earth that are inhabited by humans or are readily accessible to them.

implementing organization. The organization (and its contractors) that performs activities in order to select and investigate the suitability of a site for a nuclear facility, and that undertakes to design, construct, commission, operate and shut down such a facility.

justification, radiological. The cost-benefit basis which supports a specified decision, practice or operation that is expected to result in human exposures to radiation.

leaching, (i) Extraction of a soluble substance from a solid by a solvent withwhich the solid is in contact, (ii) The term is often used in waste management to describe the gradual erosion/dissolution of emplaced solid waste or chemicals therefrom, or the removal of sorbed material from the surface of a solid or porous bed.

licence. Formal document issued by the regulatory body for major stages in the development of a nuclear facility, defined by regulations permitting the holder (the implementing organization) to perform specified activities.

Markov chain analysis. A stochastic method of simulation analysis for analysing and solving a problem by specifying the possible states of a system, determining the probabilities for transition between those states, and using extensive mathematical formalism to derive analytical functional forms for solution.

migration retardant. A material or a natural mineral (e.g. bentonite) used in a waste repository that retards the movement of radionuclides and other chemicals away from the waste and/or repository and through the geosphere to the human environment.

models. In applied mathematics, the analytical or mathematical representation or quantification of a real system and the ways that phenomena occur within that system. Individual or sub-system models can be combined to give system models. Deterministic and probabilistic models are two types of mathematical models.

33


Monte Carlo analysis. A stochastic method of simulation analysis that involves statistical sampling techniques in obtaining a probabilistic approximation to the solution of a problem. The method requires continued sampling of

values of a large number of elementary events by the avalues of a large number of elementary events by the application of the mathematical theory of random variables.

operational period. The period during which a nuclear installation (e.g. a waste repository) is being used for its intended purpose until it is shut down and decommissioned.

optimization. As used in radiation protection practice, the process of reducing the expected detriment from human radiation exposures, through use of protective measures, to as low as reasonably achievable (ALARA), economic and social factors being taken into account.

overpack. Secondary (or additional) external containment for packaged radioactive waste.

post-sealing period. The period after a waste repository has been shut down and sealed.

probabilistic analysis. A statistical analysis technique for studying the expected behaviour of a system with parameters whose values are uncertain, with events whose occurrences are random, and with features which may or may not be present.

release scenario (see scenario analyses).

repository. The underground facility into which the waste is emplaced for disposal.

repository system. The repository and all its supporting facilities.

risk. A measure of the deleterious effects that may be expected as a result of a technology, traditionally quantified as the product of the probability and the consequence of the occurrence of an event or series of events. (In radiation protection, the term is normally used to indicate the sum of the probabilities of the deleterious effects that a given individual or population will incur as a result of a radiation dose.) However, the definition and quantification of the concept of ‘risk’ are being re-evaluated with regard to safety considerations. (See item 36 in the bibliography.)

risk analysis. An analysis of the risks associated with a technology wherein the possible events and their probabilities of occurrence are considered together with their potential consequences, the distribution of these consequences within the affected populations, and the uncertainties of these estimates.

34


safety analysis. The analysis and calculation of the hazards associated with the implementation of a proposed activity.

safety assessment. The comparison of the results of safety analyses withacceptability criteria, its evaluation, and the resultant judgements made on the acceptability of the system assessed.

scenario analyses. Safety analyses that identify and define phenomena, and their interactions, which could initiate and/or influence the release and transport of radionuclides from the source to humans. Release scenarios define the phenomena relevant to release of radionuclides from the waste sources; transport scenarios define the phenomena relevant to transport of the released radionuclides through the geosphere and biosphere to humans.

sensitivity analysis. An analysis of the variation of the solution of a problem with changes in the values of the variables involved. Two types of sensitivity analysis can be recognized: In simple parameter variation the sensitivity of the solution is investigated for changes in one or more input parameters within a reasonable range about selected reference or mean values. In the perturbation analysis the sensitivities of the solution with respect to changes in all input parameters can be obtained by applying differential and/or integral analysis.

shallow-ground disposal (e.g. shallow-ground burial). Disposal of radioactive waste, with or without engineered barriers, above or below the ground surface, where the final protective covering is of the order of a few metres thick. Some Member States consider ‘shallow-ground disposal’ to be a mode of storage rather than a mode of disposal.

shut-down and sealing. Action taken, after disposal operations have ceased, to prepare an installation for abandonment or minimum surveillance.

simulation analysis. A general method of studying the behaviour of a real system or phenomenon; the method usually involves devising a model representing the essential features of the system and carrying out the solution of the mathematical and logical relations of the model. The simulation can be either deterministic or stochastic depending on the model selected. Markov chain analysis and Monte Carlo analysis are two well known examples of stochastic simulation techniques.

stochastic event. A random event which can be predicted only by the probability of its occurrence. Regarding hydrology, for example, the term applies to data on phenomena that occur in time and/or space which are basically of a probabilistic nature, but whose values depend partially on their respective

35


time and/or space co-ordinates. In a stochastic time series, a term in the series is significantly related to the next one and this is considered in the time series analysis and synthesis.

storage. The emplacement of waste in a facility, with the intent that it will be retrieved at a later time.

tectonic. Pertaining to the rock structure and external forms resulting from the deformation of the earth’s crust.

transport scenario (see scenario analyses).

uncertainty analysis. An analysis to estimate the uncertainties and error bounds of the quantities involved in and the results from the solution of a problem. This requires the application of statistical techniques and definition of the input data information in probabilistic form.

waste form. The physical and chemical form of the waste (e.g. liquid, in concrete, in glass, etc.) without its packaging.

waste management. All activities, administrative and operational, that areinvolved in the handling, treatment, conditioning, transportation, storage and disposal of waste.

waste package. The waste form and any container(s) and other engineered barriers (e.g. absorber materials), as prepared for handling, transport, storage and/or disposal. A container may be a permanent part of the waste package or may be re-usable (e.g. shielding cask, shock absorbers, etc.) for any waste management step. The waste package may vary for the different steps in waste management.

‘worst-case’ scenario. The scenario for release and transport of radionuclides from a waste repository to the biosphere which represents the most severe situation conceivable on the basis of pessimistic assumptions. Agreement on a ‘worst-case’ scenario may be difficult. Thus, the terminology “ ‘conservative, but realistic’ scenarios” is frequently used to define a set of scenarios that can be used in sensitivity and uncertainty analyses for safety assessment purposes.

36


BIBLIOGRAPHY

1. Recommendations of the International Commission on Radiological Protection, ICRP Publication 26, Pergamon Press (1977).

2. Radionuclide Releases into the Environment: Assessment of Dose to Man, ICRP Publication 29, Pergamon Press (1979).

3. LINDELL, B., New Trends in Radiation Protection, Statens Straalskyddinstitut, Stockholm, Rep. SSI-1977-032 (1977).

4. Risk Analysis and Geologic Modelling in Relation to the Disposal o f Radioactive Waste into Geological Formations, OECD/NEA/CEC, Proc. Workshop, Ispra, Italy, May 1977.

5. Migration of Long-lived Radionuclides in the Geosphere, CEC/OECD/NEA, Proc. Seminar, Brussels, Belgium, January 1979.

6. Handling of Spent Nuclear Fuel and Final Storage of Vitrified High-Level Waste,Volume IV: Safety Analysis Kambranslesakerhet Rep., Sweden (1978).

7. “Handling and Final Storage of Unreprocessed Spent Nuclear Fuel”, Safety Analysis Karnbranslesakerhet Rep., Sweden (1978).

8. HILL, M.D., GRIMWOOD, P.D., Preliminary Assessment of the Radiological Protection Aspects of Disposal of High-level Waste in Geological Formations, Nat. Radiological Protection Board, Harwell, Rep. NRPB-R69 (1978). (See also: “Nuclear waste disposal: radiological protection aspects”, New Scientist, 11 May 1978).

9. Release Consequence Analysis for a Hypothetical Geologic Radioactive Waste Repository in Salt, INFCE/DEP/WG.7/21 (August 1979).

10. Release Consequence Analysis for a Hypothetical Geologic Radioactive Waste Repository in Hard Rock, INFCE/DEP/WG.7/21 (December 1979).

11. Status of Technologies Related to Radioactive Waste Management and Disposal, INFCE/DEP/WG.7/9 (September 1979).

12. Radiological Health and Safety Impact from the Management o f Radioactive Waste, INFCE/DEP/W G.7/15 (1979).

13. “Interdepartementale Commissie voor de Kernenergie: Rapport over de mogelijkheden van opslag van radioactieve afvalstoffen in zoutvoorkomens in Nederland” (Interdepartmental Committee for nuclear energy: Report on the possibilities for disposal of radioactive waste in salt domes in the Netherlands) (in Dutch) (April 1979).

14. BARTLETT, J.W., BURKHOLDER, H.C., WINEGARDNER, W.K., Safety Assessment of Geologic Repositories for Radioactive Waste, Battelle Pacific Northwest Labs., Richland, BNWL-SA-6068 (1977),

15. State of Geologic Knowledge Regarding Potential Transport of High-level Radioactive Waste from Deep Continental Repository, USA Environmental Protection Agency,EPA/520/4-78-004 (1978).

16. CLAIRBORNE, H.C., GERA, F., Potential Containment Failure Mechanisms and their Consequences at a Radioactive Waste Repository in Bedded Salt in New Mexico,Oak Ridge Nat. Lab., ORNL-TM-4639 (1974).

17. BURKHOLDER, H.C., CLONINGER, M.O., BAKER, D.A., JANSEN, G„ “Incentives for partitioning high-level waste”, Nucl. Technol. 31 202 (1976).

18. HAMSTRA, J., Safety Analysis for the Disposal o f Solid Radioactive Wastes in a Netherlands Salt Dome, Reactor Centrum Nederland, Petten, RCN-75-040 (April 1975).

19. HAMSTRA, J., “Radiotoxic Hazard Measure for Buried Solid Radioactive Waste”,Nucl. Saf. 16 2 (1975).

20. FRANKE, H., Betrachtungen in Havariesituationen bei der Lagerung radioaktiver Abfalle in einem Salzbergwerk, Staatliche Zentrale fur Strahlenschutz, Berlin, GDR, SZS-138 (1972).

37


21. BURKHOLDER, H.C., GREENBORG, J., STOTTLEMYRE, J.A., BRADLEY, D.J., RAYMOND, J.R., SERNE, R.J., Waste Isolation Safety Assessment Program —Technical Progress Report for FY-77, Battelle Pacific Northwest Labs., PNL-2642 (1979).

22. GREENBORG, J., et al., Scenario Analysis Methods for Use in Assessing the Safety of the Geologic Isolation of Nuclear Waste, Battelle Pacific Northwest Labs.,PNL-2643 (1979).

23. CAMPBELL, J.E., et al., Risk Methodology for Geologic Disposal of Radioactive Wastes: Interim Report, Sandia Labs., Albuquerque, SAND 78-0029 (1979).

24. GIRARDI, F., AVOGADRO, A., BERTOZZI, G., D’ALESSANDRO, M., LANZA, F., MURRAY, N.C., “A risk analysis methodology for deep underground radioactive waste repositories and related experimental research”, Underground Disposal o f Radioactive Wastes (Proc. Symp. Otaniemi, 1979) 2, IAEA, Vienna (1980) 407.

25. BRANDSTETTER, A., HARWELL, M.A., “The waste isolation safety assessment programme”, ibid. 2 423.

26. LYON, R.B., ROSINGER, E.L.J., “Safety assessment for deep underground disposal vault pathways analysis”, ibid. 2 453.

27. DEVELL, L., BERGMAN, R., BERGSTROM, U., KNELLBERT, N., STENQUIST, C., GRUNDFELT, B., “Disposal o f high-level waste or spent fuel in crystalline rock: Factors influencing calculated radiation doses”, ibid. 2 465.

28. WEBB, G.A.M., HILL, M.D., “Application of the results of radiological assessments of high-level waste disposal”, ibid. 2 509.

29. HERRENBERGER, V., SCHNEIDER, J.F., GASSMAN, J., “Site data availability and safety assessment method development for underground waste repositories”, ibid. 2 495.

30. HASTED, F., MEHLSEN, S., “Design and safety evaluation of a Danish high-level waste disposal facility in selected salt domes”, ibid. 2 539.

31. BARBREAU, A., et al., “Premieres evaluations des possibility d’evacuation des dechets radioactifs dans les roches crystallines”, ibid. 2 387.

32. BURKHOLDER, H.C., Waste Isolation Performance Assessment — A Status Report, ONWI-6O, Battelle Memorial Institute Office o f Nuclear Waste Isolation, Columbus,Ohio (1979).

33. LEVI, H.W., “The ‘Project Safety Studies Entsorgung’ in the Federal Republic of Germany”, Underground Disposal o f Radioactive Wastes (Proc. Symp. Otaniemi, 1979)2, IAEA, Vienna (1980) 437.

34. GIRAUD, G., CANDES, P., Les dechets radioactifs, Ann. Mines (Mars-Avril 1976) p. 18.35. de MARSILY, G., LEDOUX, E., BARBREAU, A., MARGAT, J., Nuclear waste disposal:

Can the geologist guarantee isolation? , Science 197 (1977) 4303.36. SCHAEFER, R.E., “What Are We Talking About When We Talk About ‘Risk’? : A

Critical Survey of Risk and Risk Preference Theories”, International Institute for Applied Systems Analysis, Laxenburg, Austria, Research Memorandum RM-78-69 (December 1978).

37. KOPLIK, C.M., et al., Status Report on Risk Assessment for Nuclear Waste Disposal, EPRI-NP-1197, Electric Power Research Institute, Palo Alto, California (October 1979).

38


DRAFTING AND REVIEWING BODIES

1. Consultants’ Meetings to prepare working draft:

IAEA, Vienna, 11 — 15 December 1978

ARGENTINA

B e n i n s o n , A.M. Comision Nacional de Energia Atomica,Avenida Libertador 8250,Buenos Aires-1429

NETHERLANDS

Glasbergen, P. Rijksinstitut voor DrinkwatervoorzieningPostbus 150, 2060 AD Leidschendam

SWEDEN

Devell, L. (C h a ir m a n ) Studsvik Energiteknik AB,Fack, S-611 82 Nykoping

UNITED KINGDOM

Shaw, K.B. National Radiation Protection Board, AERE,Harwell, Oxfordshire 0X 11 ORA

INTERNATIONAL ATOMIC ENERGY AGENCY

Richter, D.K. (S c ie n t i f i c S e c r e ta r y ) Division of Nuclear Safety and EnvironmentalProtection

IAEA, Vienna, 18—22 June 1979

SWEDEN

Devell, L. (C h a irm an ) Studsvik Energiteknik AB,Nykoping

39


Burkholder, H. Battelle Memorial Institute — ONWI,Columbus, Ohio


Richter, D.K. (S c ie n t i f i c S e c r e ta r y ) Division of Nuclear Safety and EnvironmentalProtection

2. Advisory Group Meeting, Petten, Netherlands, 8—12 October 1979

Participants

ARGENTINA

Beninson, A.M. Comision Nacional de Energia Atomica,Avenida Libertador 8250, Buenos Aires — 1429

BELGIUM

Bonne, A. CEN,Boeretang 200, B-2400 Mol

CANADA

Lyon, R.B. Whiteshell Nuclear Research Establishment, AECL,Pinawa, Manitoba ROE 1L0

CZECHOSLOVAKIA

Horyna, J. Nuclear Research Centre,250 68 Rez, near Prague

FINLAND

Vuori, S. Technical Research Centre of Finland,Nuclear Engineering Laboratory,P.O.Box 169, SF-001 80 Helsinki 18

U N IT E D S T A T E S O F A M E R IC A

40


F R A N C E

Bouville, A.

Bonnet, M.

Departement de protection,Centre d’etudes nucleaires de Fontenay-aux-Roses,B.P. No. 6, F-92260 Fontenay-aux-Roses

Bureau de recherches geologiques et minieres (BRGM), 6, rue Chasseloup Laubat, Paris 15eme

Van Kote, F. Departement de surete nucleaire,Centre d’etudes nucleaires de Fontenay-aux-Roses, F-92260 Fontenay-aux-Roses

GERMAN DEMOCRATIC REPUBLIC

Korner, W. Staatliches Amt fur Atomsicherheit und Strahlenschutz, Waldowallee 117, DDR-1157 Berlin-Karlshorst

GERMANY, FEDERAL REPUBLIC OF

Bechthold, W. Kernforschungszentrum Karlsruhe, Abt. ABRA, Postfach 3640, D-7500 Karlsruhe

INDIA

Sunder Rajan, N.S. Bhabha Atomic Research Centre, Trombay, Bombay 400 085

NETHERLANDS

Glasbergen, P. Rijksinstituut voor Drinkwatervoorziening, Postbus 150, 2060 AD Leidschendam

Baas, J.L. Ministry of Health and Environmental Protection, Dr. Reijersstraat 10—12, Leidschendam

Hamstra, J. Netherlands Energy Research Foundation ECN, Postbus 1, 1755 ZG Petten

Reij, W.C.

Verkerk, B.

Ministry of Health and Environmental Protection, Dr. Reijersstraat 1 0 -1 2 , Leidschendam

Netherlands Energy Research Foundation ECN, Postbus 1, 1 755 ZG Petten

41


SWEDEN

Devell, L. (C hairm an)

Norrby, S.

Studsvik Energiteknik AB, Fack, S-611 82 Nykoping

Statens Stralskyddinstitut, Fack, S - l04 01 Stockholm

SWITZERLAND

Herrenberg, V.

McCombie, C.

Institut federal de recherches en matiere de reacteurs, CH-5303 Wiirenlingen

Societe nationale pour l’entreposage des dechets radioactifs,

Parkstrasse 23, CH-5401 Baden

UNITED KINGDOM

Shaw, K.B. National Radiological Protection Board, Harwell, Oxfordshire 0X 11 ORA

UNITED STATES OF AMERICA

Burkholder, H. Battelle Memorial Institute — ONWI,505 King Avenue, Columbus, Ohio 43201

COMMISSION OF EUROPEAN COMMUNITIES

Cadelli, N.

Bertozzi, G.

200, rue de la Loi, B-1049 Brussels, Belgium

CCR Ispra, 1-21020 Ispra, Italy

OECD/NEA

Gera, F. 38, Boulevard Suchet, F-75016 Paris, France


Richter, D. (S c ie n t i f i c S e c r e ta r y ) Division of Nuclear Safety and Environmental Protection, IAEA,

A-1400 Vienna, Austria

42


3. Technical Review Committee on the Underground Disposal of Radioactive Waste, IAEA, Vienna, 10—14 November 1980

Committee Members

BELGIUM

Heremans, R. Geo-technologie du CEN/SCK,Boeretang 200, B-2400 Mol-Donk

CANADA

Mayman, S.A. Whiteshell Nuclear Research Establishment,Pinawa, Manitoba ROE ILO

CZECHOSLOVAKIA

Malasek, E. Czechoslovak Atomic Energy Commission, Slezska 9, 120 29 Prague 2

FRANCE

Barbreau, A. Institut de protection et de surete nucleaire, CEN de Fontenay-aux-Roses,B.P. N o .6, F-92260 Fontenay-aux-Roses

GERMAN DEMOCRATIC REPUBLIC

Runge, K. National Board of Nuclear Safety and Radiation Protection,

Waldowallee 117, DDR-115 Berlin-Karlshorst

GERMANY, FEDERAL REPUBLIC OF

Kuhn, K. Institut fur Tieflagerung der Gesellschaft fur Strahlenund Umweltforschung mbH,

Berliner Strasse 2, D-3392 Clausthal-Zellerfeld

INDIA

Sunder Rajan, N.S. High Level Waste Management Section,Eng. Hall No. 5,Bhabha Atomic Research Centre, Trombay, Bombay 400 085

43


JAPAN

Doi, K. Radioactive Waste Management Centre,No. 15 Mori Building, 4F, 2-8-10 Toranomon, Tokyo

NETHERLANDS

Baas, J.L. Ministry o f Health and Environmental Protection, P.O. Box 439, NL-2260 AK Leidschendam

SWEDEN

Larsson, A. (C ha irm an) Swedish Nuclear Power Inspectorate, Box 27106, S-102 52 Stockholm

SWITZERLAND

Rometsch, R. CEDRA,Parkstrasse 23, CH-5401 Baden

UNITED KINGDOM

Feates, F.S. Department of the Environment, Becket House, 1 Lambeth Palace Road, London SE1 7ER

UNITED STATES OF AMERICA

Vieth, D. Division of Repository Development,US Department of Energy,Mail Stop B-107 (GTN), Washington DC 20545

Experts accompanying Committee Members

AUSTRALIA

Hardy, C.J. Australian Atomic Energy Commission,Private Mail Bag, Sutherland, 2232, New South Wales

FRANCE

Barthoux, M. Commissariat a l’energie atomique,A.N.D.R.A., Agence nationalede gestion des dechets

radioactifs,31—33 rue de la Federation, F-75752 Paris 15

44


Berges, G. Secretariat du comite interministerial de la securite nucleaire,

27 rue Gudinot, F-75700 Paris

G E R M A N D E M O C R A T IC R E PU B L IC

Noack, W. National Board of Nuclear Safety and Radiation Protection,

Waldowallee 117, DDR-115 Berlin-Karlshorst

JAPAN

Moriyama, N. Reactor Safety Research Centre,Tokai Research Establishment, JAERI, Tokai-mura, Naka-gun, Ibaraki-ken

NETHERLANDS

Hamstra, J. Netherlands Energy Research Foundation, P.O. Box 1, NL-1755 ZG Petten

SWEDEN

Boge, R.

Rydell, N.

National Institute o f Radiation Protection,Box 60204, S-104 01 Stockholm

Swedish National Council for Radioactive Waste Management,

Box 5864, S - l02 48 Stockholm

UNITED KINGDOM

Grover, J.R. AERE, Harwell, Oxfordshire 0X 11 ORA

COMMISSION OF EUROPEAN COMMUNITIES

Haijtink, B. 200 rue de la Loi,B-1049 Brussels, Belgium

OECD/NEA

Gera, F. 38 boulevard Suchet, F-75016 Paris, France

45


IN T E R N A T IO N A L A TO M IC E N E R G Y A G E N C Y

Dlouhy, Z. (S c ie n t i f i c S e c r e ta r y )

Schneider, K.J. (S c ie n t i f i c S e c r e ta r y )

Richter, D.K.Irish, E.R.Heinonen, J.U.Tsyplenkov, V.

Division of Nuclear Fuel Cycle Division of Nuclear Fuel Cycle Division of Nuclear Fuel Cycle Division of Nuclear Fuel Cycle Division o f Nuclear Fuel Cycle Division of Nuclear Fuel Cycle

4. Technical editing of the final drafts:

Irish, E.R. Division of Nuclear Safety and EnvironmentalProtection,

IAEA, A-1400 Vienna, Austria

4 6


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Comision Nacional de Energfa A tom ica, Avenida del L ibertador 8250, RA-1429 Buenos AiresHunter Publications, 58 A G ippsS treet, Collingwood, V ic to ria 3066Service Courrier UNESCO, 202, Avenue du Roi, B-1060 BrusselsS .N .T .L., Spalena 51, CS-113 02 Prague 1A lfa , Publishers, Hurbanovo namestie 6, CS-893 31 BratislavaO ffice International de Docum entation et Librairie, 48, rue Gay-Lussac,F-75240 Paris Cedex 05Kultura, Hungarian Foreign Trading Company P.O. Box 149, H-1389 Budapest 62O xford Book and Stationery Co., 17, Park Street, Calcutta-700 016 O xford Book and Stationery Co., Scindia House, New D elh i-110 001 Heiliger and Co., L td ., S cientific and Medical Books, 3, Nathan Strauss Street, Jerusalem 94227L ibreria Scientifica, D ott. Lucio de Biasio "ae iou ".Via Meravigli 16, 1-20123 MilanMaruzen Company, L td ., P.O. Box 5050, 100-31 Tokyo International M artinus N ijh o ff B.V., Booksellers, Lange V oorhou t 9-11, P.O. Box 269, NL-2501 The HagueMirza Book Agency, 65, Shahrah Quaid-e-Azam, P.O. Box 729, Lahore 3 Ars Polona-Ruch, Centrala Handlu Zagranicznego,Krakowskie Przedmiescie 7, PL-00-068 Warsaw llex im , P.O. Box 136-137, BucarestVan Schaik's Bookstore (Pty) L td ., L ib ri Building, Church Street,P.O. Box 724, Pretoria 0001Diaz de Santos, Lagasca 95, Madrid-6Diaz de Santos, Balmes 417, Barcelona-6AB C.E. Fritzes Kungl. Hovbokhandel, Fredsgatan 2, P.O. Box 16356, S-103 27 StockholmHer Majesty's Stationery O ffice, Agency Section PDIB, P.O.Box 569, London SE1 9NH

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PROCEDURES AND DATA - Nucleus Safety Standards/Safety_Series_056_1981.pdf(3) Recommendations. Publications in this category, containing general recommendations on safety practices,