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34300000347132
Universiteit Vrystaat
Analysis of a Post-Closure Safety Assessment
Methodology for Radioactive Waste Disposal Systems
in South Africa
by
J.J. van Blerk
Thesis
submitted in the fulfilment of the requirements for the degree of
Doctor of Philosophy
in the Faculty of Natural Science
Department of Geohydrology
University of the Free State
Bloemfontein
February 2000
Promoter: Prof. J.F. Botha, M.Se., Ph.D. (SteIl.)
\.
ACKNOWLEGEMENTS
It is my desire to acknowledge the following organisation and persons who contributed
significantly towards finishing this thesis.
The Atomic Energy Corporation of South Africa (Ltd.) for the financial support and oppor-
tunity to write this thesis, and especially my colleagues from the Nuclear Liabilities Man-
agement Division for their support and suggestions.
My promoter, Prof. J.F. Botha from the Institute for Groundwater Studies. His confidence,
guidance and helpful suggestions, especially during the final stages of the thesis, is greatly
appreciated.
My parents, family and friends for their interest, inspiration and continuous prayers, which
are greatly appreciated.
My wife Daleen, my two sons Wihan and Derik, for their patience, patience and patience
during the completion of this thesis. There is no way in the world that I could have achieved
this. without their support and love. It is therefore with great appreciation and love that I
want to dedicate this thesis to them.
My Lord and Saviour Jesus Christ, whose presence, guidance and mercy in my life helped
me, accomplish this task.
-i-
CONTENTS
Acknowlegements i
List of Figures vii
List of Tables x
Chapter 1
Introduction
1.1 General 1
1.2 Definition of the Term Safety Assessment 3
1.3 The Situation in South Africa 5
1.4 Purpose of the Study 8
1.5 Scope of the Study 8
Chapter 2
Historical Overview and Properties of Radioactive Waste
2.1 Radioactivity ; 12
2.2 Nature and Effects of Radioactive Materials 14
2.3 Origin and Types of Radioactive Waste 16
2.3.1 General 16
2.3.2 The Nuclear Fuel Cycle 17
2.3.3 Decommissioning of Nuclear Facilities 18
2.3.4 Other Forms of Nuclear Waste 18
2.4 Classification of Radioactive Waste 19
2.4.1 General 19
2.4.2 Qualitative Classification 20
2.4.3 Quantitative Classification 21
Chapter 3
Management of Radioactive Waste
3.1 Introduction 23
3.2 An Integrated Approach to Radioactive Waste Management 23
3.2.1 General 23
3.2.2 Waste Pre-treatment 24
3.2.3 Waste Treatment and Conditioning 25
3.2.4 Organisational Responsibility 25
-ii-
3.3 Management Strategies 27
3.3.1 Fuel Recycling 27
3.3.2 Storage of Radioactive Waste 28
3.3.3 Disposal of Radioactive Waste 28
3.4 Radiation Protection Objectives 30
3.5 Waste Disposal Practices 32
3.5.1 Low- and Intermediate Level Waste 32
3.5.2 High-Level Waste 34
3.5.3 Spent Sealed Source Disposal 35
3.6 Nuclear Liabilities 36
3.6.1 Definition of Term Nuclear Liabilities 36
3.6.2 Sources of Nuclear Liabilities 37
3.6.3 Nuclear Liabilities Management 38
Chapter 4
Components of a Radioactive Waste Disposal System
4.1 Introduction 39
4.2 External Components 40
4.2.1 Repository Factors 40
4.2.2 Geological and Climatic Processes and Events 41
4.2.3 Future Human Actions and Behaviour 41
4.2.4 Other Factors 42
4.3 Internal Components 42
4.3.1 General 42
4.3.2 Site Selection 44
4.3.3 Site Characterization 45
4.3.4 Waste Characterization 45
4.3.5 The Engineered Barrier System 47
4.3.6 Human Behaviour 48
4.3.7 The Near-Field 49
4.3.8 The Geosphere 49
4.3.9 The Biosphere 51
Chapter 5
Concepts for the Disposal of Radioactive Waste
5.1 Introduction ,.. 54
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5.2 The Vaalputs National Radioactive Waste Disposal Facility 55
5.2.1 General 55
5.2.2 Site Characteristics 55
5.2.3 Waste Characteristics 57
. 5.2.4 Waste Packages 57
5.2.5 Physical Dimensions of the Waste Disposal Trenches 58
5.3 The Thabana Radioactive Waste Disposal Facility 59
5.3.1 General 59
5.3.2 Site Characteristics 61
5.3.3 Waste Characteristics 62
5.3.4 Physical Description of The Repositories 62
5.4 Concept for the Disposal of Canada's Nuclear Fuel Waste 63
5.4.1 General 63
5.4.2 Waste Characteristics 65
5.4.3 The Host Medium 66
5.4.4 Requirements for the Disposal Concept 67
5.4.5 Features of the Disposal Concept 67
5.5 The BOSS Concept for the Disposal of Spent Sources 69
5.5.1 General , ; '69
5.5.2 Requirements of The Disposal Concept 72
5.5.3 Waste Characteristics 72
5.5.4 Features of The Disposal Concept 73
5.5.5 Design of the Disposal Waste Packages 74
5.5.6 Reference Waste Packages Design for 226Ra 75
Chapter 6
Basic Principles of a Post-closure Safety Assessment
6.1 Introduction 78
6.2 Models in Safety Assessments 79
6.2.1 General 79
6.2.2 Development of A Physical Theory 80
6.2.3 The Mathematical Model.. 81
6.2.4 The Conceptual Model 83
6.2.5 Analytical and Numerical Models 84
6.3 The Nature of a Post-closure Safety Assessment 85
6.4 Purpose of a Post-closure Safety Assessment.. 85
6.5 Characteristics of the Post-closure Safety Assessment.. 87
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6.6 Framework for the Post-closure Safety Assessment.. 88
6.6.1 Safety Assessment Context 88
··6.6.2 Description of the Disposal System 91
6.6.3 Description and Justification of the Site's Future Evolution 92
6.6.4 Formulation and Implementation of Models 93
6.6.5 Consequence Analysis 94
6.6.6 Interpretation of Results 95
6.6.7 Confidence Building 96
Chapter7
Prospective Evaluation of a Disposal System
7.1 Introduction 98
7.2 The Near-Field 98
7.2.1 General 98
7.2.2 Near-Surface Disposal Systems 99
7.2.3 Geological Disposal Systems 102
7.3 Evaluation of the Geosphere 103
7.3.1 General 103
7.3.2 Adveetion 104
7.3.3 Mechanical Dispersion 104
7.3.4 Molecular Diffusion 105
7.3.5 Retardation Processes 106
7.3.6 Radioactive Decay 107
7.4 Evaluation of the Biosphere 108
ChapterS
Uncertainties in Safety Assessments
8.1 Introduction 111
8.2 Uncertainty in the Unknown Future State of the Disposal System 112
8.3 Data and Parameter Uncertainties 113
8.4 Model Uncertainty 114
8.4.1 General 114
8.4.2 Mathematical Model Uncertainty 114
8.4.3 Conceptual Model Uncertainty 115
8.4.4 Numerical and Approximation Uncertainty 116
8.4.5 Computer Code Uncertainty 116
-v-
8.5 Treatment of Future Uncertainties 117
8.5.1 General 117
8.5.2 Developing a Comprehensive List of FEPs 118
8.5.3 Screening of the FEP List 118
8.5.4 Ordering of FEPs 119
8.5.5 A Systematic Approach to Scenario Generation 120
8.5.6 Model Development for Consequence Analysis 124
8.6 Treatment of Data and Parameter Uncertainty 125
8.7 Treatment of Model Uncertainty 127
Chapter 9
A Decision Analysis Framework for Post-closure Safety Assessments
9.1 Introduction 129
9.2 Review of the Decision Analysis Framework 130
9.3 Basic Principles of A Decision Analysis 131
9.4 Decision Analysis Frameworks 132
9.4.1 The Framework of Freeze 132
9.4.2 The Disposal System Framework 133
9.4.3 The Decision Model 135
9.4.4 Data Worth Analysis and Nuclear Liabilities 138
Chapter 10
Conclusions and Recommendations
10.1 Introduction 139
10.2 Conclusions 140
10.3 Recommendations 144
Appendix A
The International ISAM List of Features, Events and Processes (FEPs) for A Radio-
active Disposal System
Al Definition of the Disposal System Domain 145
A2 Layers and Categories of the List 145
A3 The draft ISAM International FEP List (Version 1.0) in classification scheme order ..
................................................................................................................................ 146
References 150
Summary 156
Opsomming 158
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LIST OF FIGURES
Chapter 1
Introduction
Figure 1-1 Locality map of the two radioactive waste disposal facilities at Vaalputs and
Pelindaba, as well as the Nuclear Power Plant at Koeberg 6
Chapter 2
Historical Overview and Properties of Radioactive Waste
Figure 2-1 The three types of radiation emitted by radionuclides and the shielding re-
quired to stop them (after Issler, 1990) 16
Figure 2-2 Activity range for some important applications of sealed sources and the mag-
nitude of problems associated with spent sources (after IAEA, 1991) 19
Figure 2-3 The qualitative radioactive waste classification scheme proposed for the AEC.
The figures in brackets represent the estimated percentage of AEC waste (AEC,
1997a) 21
Chapter 3
Management of Radioactive Waste
Figure 3-1 Model of a typical container used for the immobilisation of low- and interme-
diate-level waste (NAGRA, 1992) 26
Figure 3-2 The different stages in the recycling of spent fuel used 27
Chapter 4
Components of a Radioactive Waste Disposal System
Figure 4-1 Conceptual representation of the different components of a disposal system,
and the flow of information between them 40
Figure 4-2 Composition of the biosphere after Torres and Simon (1997) 52
Chapter 5
Concepts for the Disposal of Radioactive Waste
Figure 5-1 Space map (Bands 4,5,3) of the Vaalputs area. The dune fields are shown in
light green, while the disposal site is shown in red 56
Figure 5-2 Geological succession at Vaalputs. (Modified from Levin, 1998) 57
- vii-
Figure 5-3 The waste packages used for the disposal of radioactive waste at Vaalputs ...
.................................................................................................................... 59
Figure 5-4 Near-surface earth trenches used for the disposal of low- and intermediate
level waste at Vaalputs 60
Figure 5-5 An excavated portion of Trench 7 at Thabana, which reflect a typical past
practice near-surface disposal concept. , 63
Figure 5-6· The engineered near-surface borehole facility (a) and stainless steel container
used for the disposal of spent 60Cosources at Thabana (b) 64
Figure 5-7 Schematic representation of the CANDU fuel bundle and container 65
Figure 5-8 The plutonic rock in the Canadian Shield has characteristics considered to be
technically favourable for the disposal of spent fuel in Canada (taken from
ABCL, 1994b) 66
Figure 5-9 Schematic representation of the natural and engineered barriers that forms
part of the geological disposal concept for the disposal of Canada's spent fuel
(taken from ABCL, 1994b) 68
Figure 5-10 Cutaway view of the disposal vault for the geological disposal facility pro-
posed for the disposal of Canada's spent fuel (AECL, 1994b) 70
Figure 5-11 Schematic representation of the various vault seals used to limit the release of
contaminants from the disposal vault of a geological disposal concept (AECL,
1994b) 71
Figure 5-12 Schematic presentation of the repository area of the BOSS disposal concept.
................................................................................................................... 73
Figure 5-13 Graphical illustration of the in situ repository configuration for the BOSS
disposal concept, showing the position of the waste packages and the use of
different materials. (Not to scale.) 74
Figure 5-14 Various stages in the preparation of waste packages proposed for the disposal
of spent sources in the BOSS disposal concept. 75
Figure 5-15 Schematic depiction of the in situ placement of waste packages with the BOSS
disposal concept. 76
Chapter 6
Basic Principles of a Post-closure Safety Assessment
Figure 6-1 Flow diagram of the ISAM approach for the post-closure safety assessment
of near-surface radioactive waste disposal systems 79
Figure 6-2 Schematic representation of the domain, n, and its boundary, an, for the para-
bolic partial differential equation in Equation (6.7) 83
Figure 6-3 The safety indicators used in the post-closure safety assessment of a radio-
- viii-
active waste disposal site and their associated safety criteria (IAEA, 1997b) .
................................................................................................................... 90
Chapter 7
Prospective Evaluation of a Disposal System
Figure 7-1 Schematic partial cross-section through the wall of a chamber in a hybrid
geological disposal system, with the different processes and components that
could influence the performance of the system's near-field. [Redrawn from
IAEA (1992).] 102
Figure 7-2 Schematic illustration of adveetion and molecular diffusion in a two-layered
heterogeneous aquifer, after Botha (1996) 106
Chapter 8
Uncertainties in Safety Assessments
Figure 8-1 Outline of the ISAM international list of features, events and processes rel-
evant for near-surface disposal systems (IAEA, 1998b) 121
Figure 8-2 A simple 2x2 interaction matrix illustrating the application of the RES matrix
method to interactions between features, events and processes in a radioactive
safety assessment. 122
Figure 8-3 A conceptual interaction and secondary interaction matrix for the Process
System approach with the components of the disposal system included ex-
plicitly .
123
Figure 8-4 Principle of a migration pathway of radionuclides through the interaction ma-
trix 124
Chapter 9
A Decision Analysis Framework for Post-closure Safety Assessments
Figure 9-1 Framework for hydro geological decision analysis as proposed by Freeze et
al. (1990) and modified by Janse van Rensburg (1992) 133
Figure 9-2 A possible decision analysis framework for the post-closure safety assess-
ment of a radioactive waste disposal site 134
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LIST OF TABLES'
Chapter 2
Historical Overview and Properties of Radioactive Waste
Table 2-1 A brief chronicle of the discovery and applications of radioactivity between
1896 and 1992 (Issler, 1990; USDOE, 1998; IAEA, 1991) 13
Table 2-2 Important properties of radioactive waste used as criteria to classify them
(!AEA, 1994a) 20
Chapter 4
Components of a Radioactive Waste Disposal System
Table 4-1 Examples of long-term geological and climatic factors that may influence the
safety of a disposal site adversely 43
Table 4-2 Examples of future human actions that can affect the post-closure safety as-
sessment of a radioactive disposal facility adversely and therefore should be
included in a regional site characterization 43
Table 4-3 Examples of geological and surface factors that should be included in a re-
gional site characterization 44
Table 4-4 Information needed for a safety assessment analysis of 90Sr and 60Co(Seitz et
al., 1992) 46
Table 4-5 Examples of human characteristics that should be included for the.critical group
in the assessment of a radioactive waste disposal site 49
Table 4-6 Partial list of factors that are important in characterizing the geosphere of a
disposal site 50
Table 4-7 Summary of physical processes that may cause contamination of the terres-
trial food chain by nuclides that leaked from a radioactive waste repository.
[After Torres and Simon (1997).] 53
Chapter 5
Concepts for the Disposal of Radioactive Waste
Table 5-1 The low- and intermediate level waste inventory at Vaalputs in May 1998 ...
.................................................................................................................... 58
Table 5-2 A typical inventory of spent sources for which the BOSS disposal concept is
implied. (All dimensions refer to the height and diameter of a cylinder.) .. 72
-x-
Chapter 7
Prospective Evaluation of a Disposal System
Table 7-1 Degradation processes in engineered barriers associated with the infiltration of
water, taken from Andrade (1997) 100
Table 7-2 Parameters and equations often used to compute the radiation dose received by
humans through pathways in the biosphere of a radioactive waste disposal site .
.................................................................................................................... 109
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CHAPTER 1
INTRODUCTION
1.1 GENERAL
All users of radioactive materials in the nuclear industry, agriculture, research and medicine
generate waste, much in the same way as other human activities. The largest volumes of
radioactive waste, however, are produced by activities related to the nuclear fuel cycle. In
this cycle, radioactive waste is a by-product from the mining and milling of uranium ores,
the uranium enrichment and conversion process, and nuclear fuel fabrication. In addition to
the nuclear fuel cycle, waste generated from the operation of nuclear reactors, reprocessing
of spent fuel, and decommissioning of nuclear facilities contribute considerably to the total
volume of radioactive waste.
Radioactive waste, like all other waste, does not have any economic value and must ulti-
mately be disposed. Although radioactivity is a natural phenomenon that decreases expo-
nentially with time, radioactive waste can pose considerable risks to the natural environ-
ment of the earth. Special precautions must therefore be taken with the waste, from its
creation to its disposal-even to the time the activity has decayed to the natural level-by
public demand. This is clearly a management problem, hence the name radioactive waste
management. In its broadest sense, the term includes the activities of the waste generators,
the operators (managers) of waste disposal facilities and regulatory authorities if the waste
is subject to regulatory control.
During the first decade of the nuclear era, scientists and licensing authorities essentially
handled the problem of radioactive waste management, since they were directly confronted
with the necessity to develop reasonable and safe solutions. After decades of research, study
and testing, there exists today a large number of technological solutions with which radio-
active waste can be managed safely (IAEA, 1992). However, radioactive waste manage-
ment and waste disposal, in particular, are no longer matters for scientists alone, but re-
quires the co-operation of scientists, politicians, licensing authorities, industry and the pub-
lic at large. The international community has consequently developed a set of principles,
summarized in Table 1-1, through the International Atomic Energy Agency (!AEA) with
the main objective:
'to deal with radioactive waste in a manner that protects human health and the
-1-
Table 1-1 Fundamental principles agreed upon by the international community for the management of radioactive waste (IAEA, 1995a).
Principle Description
Radioactive waste shall be managed in such a way as to secure an acceptable level of protection to human
Protection of Human Health
health.
Radioactive waste shall be managed in such a way as to secure an acceptable level of protection to the
Protection of the Environment
environment.
Radioactive waste shall be managed in such a way as to assure that possible effects on human health and
Protection beyond National Borders ....
the environment beyond national borders will be taken into account. a..,
Radioactive waste shall be managed in such a way that predicted impacts on the health of future oQ,
Protection of Future Generations c
generations will not be greater than the relevant levels of impact that are acceptable today.
ë-~ï
Radioactive waste shall be managed in such a way that it will not impose undue burdens on future :I
Burdens on Future Generations
generations.
Radioactive waste shall be managed within an appropriate national legal framework including clear
National Legal Framework
allocation of responsibilities and provision for independent regulatory functions.
Control of Radioactive Waste Generation Generation of radioactive waste shall be kept to the minimurn practicable.
Radioactive Waste Generation and Management Interdependencies among all steps in radioactive waste generation and management shall be appropriately
Interdependencies taken into account.
The safety of the facilities for radioactive waste management shall be appropriately assured during their
Safety of Facilities
lifetime.
-
N
Introduction 3
environment, now and in the future, without imposing undue burdens on future gen-
erations.' IAEA (1995a).
However, it is clear that these principles can only be implemented successfully if member
countries create their own national legal frameworks, and associated organizational struc-
tures, to dispose and manage their radioactive wastes appropriately (IAEA, 1995a).
The majority of countries in the world today require that the long-term safety of any hazard-
ous waste disposal site must be demonstrated convincingly prior to its implementation.
However, to achieve this one must have a framework, also known as the disposal system,
that describes the disposal method and site in detail (AECL, 1994a). This includes the
potentially affected geology and accessible environment (e.g. air, land, water, people, plant
and animal life) surrounding the site. This procedure is commonly referred to as a safety or
performance assessment of the site, which will now be described in more detail.
1.2 DEFINITION OF THE TERM SAFETYASSESSMENT
According to Cho et al. (1990), the general objective of a safety assessment should be to
determine what impact the disposed waste would have on individuals and their environment
as a function of time. This implies that one must determine how radioactive materials may
escape from the disposal site and along which paths can it migrate and what effect it will ulti-
mately have on human beings. This exercise is generally referred to as a safety assessment.
The main idea behind a safety assessment is clearly to investigate, quantify and explain the
effects that a proposed (or selected) radioactive waste disposal system will have on its sur-
roundings. Although considerable attention has been given to this problem over the years,
there does not seem to be a universal view of what should be done in this regard, how it
should be done and for what reason.
The fundamental principles for radioactive waste management in Table 1-1 can be used to
divide the time during which a radioactive waste disposal site will remain active into two
periods.
(a) The pre-closure period - The time from the moment the site was developed until it
was closed and perhaps a number of years after that. This period is sometimes also
referred to as the operational period.
(b) The post-closure period - This period should theoretically begin the moment the
site is closed and extend to the time that the waste does not pose any further thread
to humanity and the environment. For this purpose, the post-closure period can be
Introduction 4
further divided into institutional and post-institutional control periods. However, the
emphasis in the fundamental principles of Table I-I, is clearly more centred on later
times (cf. the reference to future generations). The term will consequently be inter-
preted here as an unspecified period beginning at a time when nobody is interested
in maintaining the site, or know of the site.
This division suggests that the safety assessment of the site can also be divided into two
phases-a pre-closure safety assessment and a post-closure safety assessment. Since there
will be people that operate and manage the site during the pre-closure period, but not neces-
sarily during the post-closure period, a pre-closure assessment may differ considerably from
a post-closure assessment. For example, the operators of the site will certainly be able to
monitor and investigate the site and take corrective actions where necessary. Risks to hu-
mans during the pre-closure period are also different than during the post-closure period.
For example, the risk is zero that a container will drop on an operator or that workers will be
exposed directly to radiation after the site is closed. A pre-elesure assessment of the site
could therefore be based on sound scientific principles and procedures, but not a post-clo-
sure assessment. In this case, one will have to rely mainly on assumptions and guesses of
what the situation will be in the post-closure period. Since a considerable amount of work
has already been done on the pre-closure safety assessment of radioactive waste disposal
sites this thesis will concentrate exclusively on the post-closure safety assessment. How-
ever, this does not suggest that the two periods should be treated independently. In fact, a
major disadvantage of previous safety assessment analyses in South Africa is that no atten-
tion was paid to the influence that pre-elosure and operational activities might have on a
post-closure safety assessments of the sites.
The IAEA (1993a) defines performance assessment formally as:
'...an analysis to predict the performance of a disposal system, or sub-system, fol-
lowed by a comparison of the results of such an analysis with appropriate standards
or criteria.'
A performance assessment becomes a safety assessment when
'... the system under consideration is the waste disposal system and the performance
measure is the radiological impact or some other global measure of impact on the
safety of humans and the environment.'
The Scientific Committee 87-3 established by the National Council on Radiation Protection
(NRCP) used these definitions to establish suitable guidelines and concepts for conducting
a post-closure safety assessment for radioactive waste disposal facilities. In their findings
Introduction 5
(Kennedy, 1997), the committee portrayed a post-closure safety assessment as a multi-dis-
ciplinary, iterative process focussed on regulatory compliance rather than an analysis of a
disposal system for the purpose of predicting its actual behaviour. With this in mind, they
defined a post-closure safety assessment as: 'the iterative process involving site-specific,
prospective evaluations of the post-closure phase of the system' with three primary objec-
tives:
(a) determine whether reasonable assurance of compliance with quantitative perform-
ance objectives can be demonstrated,
(b) identify data, design and other needs to reach defensible decisions about regulatory
compliance, and
(c) identify waste acceptance criteria-i.e. a list of the waste types for which the reposi-
tory is intended-related to the quantities of wastes that need to be disposed.
1.3 THE SITUATION IN SOUTH AFRICA
Radioactive waste in South Africa is generated mainly by two agencies: the Atomic Energy
Corporation of South Africa Ltd. (AEC) at Pelindaba, and the Electricity Supply Commis-
sion ESKOM. The AEC is involved in many of the activities related to the nuclear fuel
cycle, while ESKOM is responsible for the operation of the Koeberg Nuclear Power Plant
near Cape Town in the Western Cape Province, as indicated in Figure 1-1. Users of radioiso-
topes in industry, medicine and research, the mines and mineral processing plants also pro-
duce various types of radioactive wastes.
Two sites are currently used for the disposal of radioactive waste in South Africa. The first
site, Thabana (previously known as Radiation Hill), is situated at Pelindaba near Pretoria in
the North-West Province (see Figure 1-1). This site has been in operation since 1969 and
consists of a variety of earth trenches used for the disposal of uranium-contaminated waste
and some plutonium. A stainless steel engineered borehole is also used for the disposal of
60COsources. The second site is the National Radioactive Waste Disposal Facility at Vaalputs
near Springbok in the Northern Cape Province (see Figure 1-1). This site came into opera-
tion in 1986 and is currently being used for the disposal of low- and intermediate level
waste from Koeberg in near-surface trenches.
The Council for Nuclear Safety (CNS), a statuary body established by the Nuclear Safety
Act of 1993 and earlier acts control the disposal of nuclear waste in South Africa exclu-
sively. This body has already granted a license for the disposal of radioactive waste atVaalputs
in 1986, after the site went through a detailed screening, selection and characterization
.-
PROVINCE
?' r/ietersburg
-....:IÓ
Q.
Cn
o-·
:I
Atlantic Indian
Ocean EASTERN CAPE Ocean
Bisho@
East London
Figure 1-1 Locality map of the two radioactive waste disposal facilities at Vaalputs and Pelindaba, as well as the Nuclear Power Plant at
Koeberg.
0\
Introduction 7
process (Corner and Scott, 1980; Moore et al., 1987). However, very little is known about
the selection criteria used for issuing a license to permanently store active solid waste at
Thabana by the then Safety Committee of the Atomic Energy Board (Niebuhr et al., 1968).
One difficulty with the current Vaalputs license is that it does not address the post-closure
safety of the facility adequately. The methods used to demonstrate the safety characteristics
of the site also did not conform to the current internationally recognized standards for per-
forming the post-closure safety assessment of a radioactive waste disposal facility. This
view was confirmed by an Expert Mission of the International Atomic Energy Agency (IAEA)
to Vaalputs (IAEA, 1998a). They concluded that, although there do exist documents that
stipulate procedures to ensure the operational safety of the site, and to protect the environ-
ment in the short term, very little has been done to ensure its post-closure safety. Particular
points raised by them include the following:
(a) The isolation strategy-the safety principles on which waste disposal at Vaalputs is
based-is not documented adequately.
(b) The waste acceptance criteria do not specify the nuclides that can be accepted for
disposal.
(c) The nature of the containment barriers and their function in achieving waste isola-
tion is not documented adequately.
(d) No provision was made for a policy to control the disposal site and nearby areas in
the post-closure period.
The strategies followed to obtain licenses for Vaalputs and Thabana were probably appro-
priate and sufficient at the times they were granted, but not today. One particular disadvan-
tage of the present management systems is that they fail to address the consequences of
today's actions on future generations. This applies especially to spent fuel produced by the
reactors at Pelindaba and Koeberg and the disposal practices at Thabana. All the spent fuel
at Koeberg is currently stored in reactor pools, while some are stored in reactor pools and
the rest in dry storage at Pelindaba. Since a permanent solution for this problem has not
receive much attention in South Africa, the country can be accused of leaving the problem
to future generations to solve (Hambleton-Jones et al., 1998). This violates one of the fun-
damental principles of radioactive waste management discussed above.
As far as Thabana is concerned, there is confusion whether it should be considered as a
storage or a disposal facility. The license granted to the AEC in 1969 was for the permanent
storage of radioactive waste without the intention of retrieval (Niebuhr et al., 1968). The
AEC has not obtained a license from the CNS for the disposal of radioactive waste at the
Introduction 8
site, and no assessment of the site's suitability as a disposal site has ever been undertaken. It
is therefore not certain whether the site is suitable for the disposal of radioactive waste.
Moreover, it is very difficult to determine what long-term consequences the past and current
practices have on humans and the environment, since there is very little information on the
practices used previously to dispose radioactive waste at Thabana. Considerable uncertain-
ties therefore exist about the total inventory and its future, that is; should the site be up-
graded to an approved disposal facility or rehabilitated and closed. In the meantime, Thabana
has to be considered as a disposal facility, for it will be necessary to quantify the current
inventory of the waste, its movement through the disposal system and its potential impact on
future generations, and to undertake the necessary remedial actions if an when necessary.
1.4 PURPOSE OF THE STUDY
The previously described situations call for the definition and implementation of a post-
closure safety assessment strategy for radioactive waste disposal facilities in South Africa,
based on the well-established and accepted international methodology. However, there does
not exist, at least at this moment, a document that describes the methodology, its implica-
tions and the steps necessary to implement it in South Africa. In this thesis an attempt is
made to develop such a structured methodology for radioactive disposal sites in South Af-
rica and other parts of Africa, in such a way that it can also be understood by interested
members of the public.
The aim of the thesis is not to present new priciples of radioactive waste management,
except for a new management strategy to dispose spent nuclear sources in boreholes, but
rather to use the existing information to develop such a structured methodology. This could
only be achieved by using the information from various organizations, regulatory authori-
ties and other interested parties. Many of the principles discussed in this thesis are therefore
not new. However, it is believed that the structured approach advanced here-the integra-
tion of radioactive waste management, post-closure safety assessment of radioactive waste
disposal sytems and the associated nuclear liabilities-is original.
A post-closure safety assessment of radioactive waste disposal systems is an extensive exer-
cise that requires input from various scientific, engineering, social and economic disci-
plines, which cannot be adequately covered by an individual. The thesis therefore concen-
trates exclusively on the broad principles of such a post-closure assessment and not its
detailed technical implementation. Readers interested in the more technical aspects are re-
ferred to (Kozak, et al., 1999), where the practical implementation of a preliminary version
of the methodology is described in detail.
Introduction 9
1.5 SCOPE OF THE STUDY
Radioactive waste management is the concluding part of all activities related to the use of
radioactive materials in the nuclear industry, agriculture, research and medicine. Tradition-
ally, the management and other activities related to the pre-closure phase of a waste dis-
posal site were considered separately from any post-closure assessment of the impact that
these activities may have on humans and their environment in the future. However, the
discussion in Section 1.2 shows that this is not desirable, at least as far as radioactive waste
is concerned, and that the two periods should be considered simultaneously within an inte-
grated framework for the management of radioactive waste.
It is obvious that one must have a detailed knowledge of all the various factors that may
influence the disposal, before such a framework can be designed. In the case of radioactive
waste these factors include the nature and properties of radioactive materials, the origin and
types of radioactive waste and the method that will be used to dispose the waste. The effects
that waste have on humans and the natural environment will vary with the process that
generates it and how it is treated before disposal. It is therefore advantageous to have a
scheme which can be used to classify the waste, albeit qualitatively. The scheme commonly
used in the nuclear industry is discussed in Chapter 2, together with the nature and proper-
ties of radioactive materials and the origin and types of radioactive waste.
In the early days, radioactive waste were often stored for an interim period to allow for
some decay, and then disposed of by dispersion and dilution in natural reservoirs (IAEA,
1993b). However, the production of larger quantities (and more dangerous) of radioactive
waste made it necessary to develop new management strategies that take advantage of the
properties and nature of radioactive waste, discussed in Chapter 2. Some of the more impor-
tant strategies that have been advanced over the years for the management of radioactive
waste and their relation to the integrated management of radioactive waste are discussed in
Chapter 3. Although there are a number of strategies that can be followed the discussion
shows that disposal is the only suitable method for the long-term management of radioac-
tive waste. However, to use this approach one must ensure that the disposed waste wil not
affect humans and their environment adversely. A number of the proposed practices, their
associated management practices and so-called nuclear liabilities are also discussed.
A disposal strategy can only be implemented if it can be demonstrated that it satisfies the
principles of radioactive waste management. This objective is best achieved by performing
an integrated post-closure safety assessment of the disposal system. However, this can only
be achieved if one has a good knowledge of the components of the system and how they are
Introduction 10
incorporated in a safety assessment, as described in Chapter 4. These components, which
can be conveniently divided into internal and external components, provide the necessary
information to characterise the movement of radionuclides through what is called the near
field, geosphere and biosphere of the system.
A few disposal concepts for radioactive waste that are of particular interest to South Africa
are discussed in Chapter 5. This includes a concept for the disposal of low- and intermedi-
ate-level waste at Vaalputs and one for the disposal of various categories of waste at Thabana.
A permanent solution for the high-level waste generated by Koeberg and the AEC has not
received much attention in South Africa. In fact, not even a conceptual plan of the disposal
concept that will be used in South Africa has yet been formulated. The conceptual geologi-
cal disposal concept prepared by Atomic Energy of Canada Ltd. for the disposal of spent
fuel in Canada is consequently used as an example of what can be expected, if one wants to
implement such a concept. The main reason for choosing this site is that its geological
characteristics are very similar to that at Vaalputs, where such a facility may likely be im-
plemented. The discussion concludes with the description of the borehole concept, recently
proposed for the disposal of spent nuclear sources in South Africa and other African coun-
tries. These examples show that design of a disposal concept will very much depend on the
.properties and characteristics of the waste that is to be disposed.
An attempt is made in Chapter 6 to clarify the often misuse of the term model in safety
assessments, before discussing a broader perspective of the basic principles of a post-clo-
sure safety assessment. The reason for this is that models play a very important role in the
safety assessment methodology. The proposed methodology, which is based on various
guidelines already accepted or considered by the international community, exhibits some
unique characteristics, when compared with similar methodologies for the disposal of other
hazardous waste.
The assessment of a radioactive waste disposal site was historically often seen as an attempt
to predict the behaviour of the site far into the future with deterministic, predictive models.
These models are based on the physical principles that underlie a specific phenomenon and
should therefore be able to predict the future behaviour of the phenomenon. Unfortunately,
this requires information on parameters whose behaviour cannot be determined with cer-
tainty far into the future. It is therefore impossible to apply the models in the historical
sense of a safety assessment. The purpose of the heuristic and phenomenological models
discussed in Chapter 7 for the evaluation and migration of radionuclides through the near
field, geosphere and biosphere, is thus not to predict the future, but rather as an aid to assess
whether a site will comply with regulatory or community imposed safety constraints. How-
Introduction 11
ever, it is important to note that a safety assessment is not an exact procedure. It will there-
fore be impossible to demonstrate complete compliance with the imposed constrains, even
with the technological aids available today. Nevertheless, it is believed that by using the
assessment methodology proposed here, it will be possible to demonstrate the safety of a
site with reasonable assurance.
A major contributor to the inexact nature of a safety assessment is the uncertainties that are
inherently part of the analysis, as discussed in Chapter 8. These uncertainties can be con-
veniently divided into uncertainties related to the unknown future state of the disposal sys-
tem, data and parameter uncertainties, and model uncertainties. The chapter is concluded
with a discussion on how to treat these uncertainties in a post-closure safety assessment.
The safety assessment methodology developed herein can be described as a decision tool to
determine the conditions under which compliance with safety objectives can be reasonably
assured and consequently very much resembles system analysis. However, no attempt has
been made in the past to formally include system analysis theory into the post-closure safety
assessment of radioactive waste disposal systems. It is also not the purpose of this thesis to
do so either. What is discussed, in Chapter 9, is a decision analysis framework that will aid
in making more reliable decisions in such an assessment and also reduce nuclear liabilities.
CHAPTER2
HISTORICAL OVERVIEW AND PROPERTIES OF RADIOACTIVE
WASTE
2.1 RADIOACTIVITY
The discovery of radioactivity dates back to 1896 when the French physicist Henri Bec-
querel tried to relate the emanations from the fluorescent mineral pitchblende to the X-rays,
discovered in 1895 by the German physicist W.C. Rontgen. Between 1896 and 1898, Bec-
querel and his students Pierre and Marie Curie discovered various naturally occurring ra-
dioactive elements such as uranium, radium and thorium, as indicated in Table 2-1. Today,
the application of radioactive materials range from medical, industrial and research tech-
niques, to the generation of nuclear energy and the manufacturing of nuclear weapons.
Despite the often tragic consequences of errors made in handling and applying radioactive
materials during the following decades, the general fascination with radioactivity did not
abate in any way. Even as late as the nineteen-thirties, charlatans were promoting the use of
radioactive toothpaste, radium hair-tonic and salve and cloth impregnated with radium.
Mineral water with a high radon content was also frequently prescribed as being good for
the health (IAEA, 1991). However, advances in both the theoretical principles and practical
applications of radioactivity have since led to the introduction of radiation protection regu-
lations that assure high levels of safety if applied correctly.
The discussion on the properties of radioactive waste will start with a review of the nature
and effects of radioactive material in Section 2.2, as discussed by Chapman and McKinley
(1988). This discussion will provide greater clarity on the principles and methodologies
applied in the management of waste generated from nuclear related activities.
Radioactive waste is generated from a variety of sources and depending on the origin and
type of radioactive waste, will exhibit different properties. Section 2.3 is consequently de-
voted to a discussion on the origin and types of radioactive waste generated from various
sources. The different properties of radioactive waste suggest that not all waste should be
treated the same, but instead to define categories of waste with similar properties that can be
managed accordingly. This led to different classification schemes for radioactive waste that
will be discussed in Section 2.4.
-12-
Table 2-1 A brief chronicle of the discovery and applications of radioactiv ity between 1896 and 1992 (Issler, 1990; USDOE, 1998; !AEA, 1991).
1896 Becquerel discovers that invisible emanations from the native oxidee uranium, pitchblende, affected photographic plates and called
them radioactive rays.
1898 Marie and Pierre Curie isolate radium from pitchblende, which was much more effective in darkening photographic plates than
pitchblende itself, and later the radioactive elements actinium, polornium and thorium.
1899 Rutherford shows that there are three types of radioactive rays, whic h he called a-, /3- and ')I-rays, names that persist today.
1901 Marie Curie provided a physician at a Paris hospital with a radiurmr source to be used for medical treatment. The source was to be
applied to a malignant surface tumour. Two years later the first SUCCIessful treatment was reported. ::t:Vi·
1904 The first attempt to treat a tumour inside the body was made by inse rting a glass capsule containing radium into the patient. .-o,
Rutherford's model of the atom distinguishes between the nucleus and the electron cloud of an atom. Georg von Hevesy conceives ;:;.
1911 the idea of radioactive tracers. The idea is later applied to medical diagnosis and the transport of radioactivity in ground water. ao
1927 Herman Blumgart, a Boston physician, first uses radioactive tracers t<o diagnose heart disease. .t-t<,
1938 Two German scientists, Otto Hahn and Fritz Strassmann discover that bombarding uranium atoms with neutrons produces barium -<
atoms. Lize Meitner explains this in terms of nuclear fission. it·~
1939 Enrico Fermi discovers the possibility of a chain reaction. Hans Bellhe identifies nuclear fusion as the energy of the sun. III
::I
1942 Fermi demonstrates the first self-sustaining nuclear chain reaction in the Chicago reactor. C.
1944 The first nuclear reactor begins operation in the USA and the first disposal in Oak Ridge, Tennessee. .~,o
1945 The USA explodes the first atomic device at a site near Alamogort da in New Mexico on the 24
11l of July, and drops the first atomic "t.0t,
bomb on the 6th of August on Hiroshima and the second on the 16th cof August on Nagasaki. ;-;.
1951 The first usable electricity from nuclear fission is produced. (IJo
1955 Arco, Idaho becomes the first U.S. town to be powered by nuclear elnergy. ...,~
1957 Radiation is released when the graphite core of the Windscale Nutclear Reactor in England catches fire. The International Atomic
III
C.
Energy Agency (IAEA) is formed to promote the peaceful uses of fiul clear energy. o'
III
1963 The United States and Soviet Union sign the Limited Test Ban TIreaty, which prohibits underwater, atmospheric and outer space ~
nuclear tests. <'tt
1966 The large number of utility orders for nuclear power reactors makes nuclear power a commercial reality in the USA. :;;
1968 I\lThe Nuclear Non-proliferation Treaty-calling for halting the spreacd of nuclear weapons capabilities-is signed. (IJ
1972 Computer axial tomography, commonly known as CAT scanning, is introduced. -tt
1977 United States president, Jimmy Carter, bans the recycling of used nuclear fuel from commercial reactors.
1979 The Three Mile Island nuclear power plant near Harrisburg, Penrnsylvania suffers a partial core meltdown. Minimal radioactive
material is released.
1986 The Chernobyl Nuclear Reactor melts down in the Soviet Union . Massive quantities of radioactive material are released in the
atmosphere over Europe.
1987 Yucca Mountain, Nevada, is designated as the prime site for the first geological repository in the United States. ....
1992 One hundred and ten commercial nuclear reactors are 0 erating in the United States. t..I
Historical Overview and Properties of Radioactive Waste 14
2.2 NATURE AND EFFECTS OF RADIOACTIVE MATERIALS
Matter is composed of atoms. The nucleus of such an atom is composed of positively charged
protons and neutral neutrons. However, an atom is electrically neutral, since the positive
charge of the protons is balanced by the negative charges of the identical number of elec-
trons, which surround the nucleus. These electrons are responsible for the chemical proper-
ties of the atom.
All atoms with the same number of protons are chemically identical and called an element.
The number of protons in a nucleus (atomic number) is consequently often represented by
the chemical symbol for the element, e.g. U for uranium, or H for hydrogen. However,
atoms of the same element sometimes contain different numbers of neutrons. These atoms
are known as isotopes and denoted by the chemical symbol of the element superscripted to
the left with the sum of its protons and neutrons (mass number), e.g. 238U. Situations often
arise though where it is necessary to describe a specific atom of an element, or nuclide,
more precisely. In such cases the number of protons is added as a subscript to the left of the
238U
symbol and the number of neutrons as a subscript to the right of the symbol, e.g., 92 146.
The majority of the known nuclides are inherently unstable and transform spontaneously
into other nuclides by emitting an a-, af3, or a y-ray, or split into two or more nuclides-a
process called fission. These processes are called radioactive decay, hence the name
radionuclides for nuclides that exhibit one or more of these processes spontaneously. The
a- and f3-rays emitted by radionuclides are particles, the nucleus of the 4He nuclide and
positively and negatively charged electrons (e±) respectively, while y-rays are, like light,
electromagnetic radiation with a very short wavelength.
The radionuclide that decays first is commonly referred to as the mother or parent nuclide,
and the product nuclide as the daughter. Many daughter nuclides are often also unstable and
decay again, thereby creating what is known as a decay chain.
Numerous experiments have shown that the rate at which a sample of radionuclides disinte-
grates is directly proportional to the number of atoms, Nit), in the sample at the time t, that is
DN(t) = -AN(t) (2.1)
where A is known as the decay constant, that varies from isotope to isotope. This property is
conventionally characterized by another constant, known as the half-life of the radionuclide
and denoted by the symbol tin. The half-life is, as the name implies, the time required for a
Historical Overview and Properties of Radioactive Waste 15
large number of atoms of a particular radionuclide to decrease by 50%, in other words from
their original number, No at t = 0, to Ni2 at t = tin' The half-life should therefore be related
to the decay constant, as it is indeed. To show this, one merely has to integrate Equation
(2.1) over time to obtain
N(t) = No exp( -At)
and then replace t by 0 and tin and N with No and NJ2 respectively, which yields
The half-lives of the radionuclides known today vary from a fraction of a second to millions
of years. For example the half-life of 219Pais 5.3.10-8 s, while that of 128Teis 8.0.1024years.
The activity or rate at which a radioactive material decays, is measured in terms of the
number of nuclei which decay or disintegrate each second. The SI-unit for this quantity is
the becquerel (Bq), defined as a decay rate of 1 nucleus per second. The unit, unfortunately,
says nothing about the type of radiation, its energy, or its ability to interact with matter. This
led.to the introduction of the term absorbed dose, defined as the energy absorbed by a body
exposed to radioactive radiation, with SI-unit the Gray (Gy), defined as the energy absorbed
by 1kg of the material.
Experience over the years has shown that the effect radioactive radiation has on an indi-
vidual differs from organ to organ and also with the individual's age. This property is de-
scribed by the term equivalent dose, obtained by multiplying the dose for a particular organ
with a suitable factor, that depends on the biological effect the type of incident radiation
may have on the organ. However, one is usually not so much interested in damage to a
particular organ in the management of radioactive waste, but rather to the individual as
such. The effective dose, defined as the sum of the dose equivalents for each part of the
human body, weighted by factors that take the susceptibilities of the different organs to
radiation-induced damage into account, is such a quantity. The sievert (Sv) is the SI-unit for
both the equivalent dose and the effective dose.
There are essentially three ways in which an individual may be exposed to radioactive ra-
diation. The first two are the inhalation of radionuclides in the form of gasses or airborne
particles, and the ingestion of radioactive foods or drinks (especially water), as is the case
with other hazardous substances. The third one, damage caused by exposing the body di-
rectly to the radiation, however, is unique to radionuclides.
Historical Overview and Properties of Radioactive Waste 16
The main reason why radioactive radiation is biologically harmful, is its ability to split and
kill individual cells, or prevents them from reproducing normally, thereby causing the de-
velopment of malignant tumours and other medical abnormalities. This ability depends in
the first place on the strength with which the different rays interact with matter. As illus-
trated in Figure 2-1, a-particles interact very strongly with matter and are easily topped by
a sheet of paper, while ,B-particles (which interact less strongly with matter), can be stopped
by a thin layer of metal. The y-particles, however, are very penetrating and can only be
stopped by thick shielding materials, such as lead or concrete. Elements that emit,B- and y-
rays are therefore the main sources of concern in the management of radioactive waste,
particularly in those cases where the waste poses an exposure risk to the public. This does
not mean though that a-emitting nuclides are not important in radioactive waste manage-
ment. On the contrary, these nuclides often cause the greatest damage to internal organs
when inhaled or ingested by an individual.
Alpha rays
• Emission of a-particles
(Helium nuclei)
Beta rays
Emission of .8-partlcles
(Electrons) Sheet of Several mm of < I mof Several metres
Paper aluminium lead of concrete
Gamma rays
Emission of
Electro-magnetic waves
Figure 2-1 The three types of radiation emitted by radionuclides and the shielding re-
quired to stop them (after Issler, 1990).
2.3 ORIGIN AND TYPES OF RADIOACTIVE WASTE
2.3.1 General
As mentioned in Chapter 1, radioactive waste is generated mainly by processes related to
the nuclear fuel cycle, the operation of nuclear reactors and the decommissioning of nuclear
facilities, with smaller quantities produced by industry, research institutions and the medi-
cal profession. The different origins of radioactive waste can cause a variation in its physi-
cal state (solid, liquid or gaseous), activity and type. Activity levels range from extremely
Historical Overview and Properties of Radioactive Waste 17
high for spent fuel, to very low for radioisotope applications in laboratories, hospitals and
universities. Equally broad is the spectrum of the half-lives of radionuclides contained in
the waste. Since the waste generated by one process may pose a different risk to humans
and the environment than that generated by another, it may be worthwhile to briefly review
these processes before continuing this discussion on radioactive waste.
2.3.2 The Nuclear Fuel Cycle
The term nuclear fuel cycle is used to denote all processes connected with nuclear power
generation, including the mining and milling of fissile materials, enrichment, production,
utilisation and storage of nuclear fuel, optional reprocessing of spent fuel, and the process-
ing and disposal of the resulting wastes (IAEA, 1993a). There are essentially three proc-
esses in the nuclear fuel cycle where waste is generated (IAEA, 1992): the mining, milling
and refinement of uranium ore, the production (enrichment, or conversion) of reactor fuel
elements, and the production of spent fuel in nuclear reactors.
At the front end of the fuel cycle, uranium is mined, milled, chemically processed and
usually enriched in the isotope 235U to a concentration of between 3 and 3.5%, to produce
nuclear fuel. In the enrichment process the solid uranium oxide (U 0 ) is first converted
3 8
chemically to the gaseous uranium hexaflouride (UF), by a process called fluorination, and
the enriched fraction of UF6 converted to uranium dioxide (UO). This is then used to pro-
duce the fuel elements for nuclear reactors. Small amounts of liquid and solid wastes, con-
taminated with uranium, are produced during this process. The calcium fluoride, generated
by the conversion of UF6 to U02' is especially important in the regard.
The radioactive waste generated by the mining and milling of uranium and thorium ores,
are usually deposited on tailings dams at the mines. These materials contain relatively low
concentrations oflong-lived radionuclides, of which uranium's daughter elements, thorium,
radium and radon are the most important.
The waste generated by a nuclear reactor can be broadly divided into two classes, the solid
and solidified waste, containing corrosion, activation and fission products, and the spent
fuel. The first class arises from the cleaning of the reactors cooling systems, fuel storage
ponds and the decontamination of equipment. This waste consists mainly of paper, filters,
ion exchangers, contaminated clothing and equipment, floor sweepings and concrete.
The spent fuel is sometimes reprocessed to regain various "isotopes, particularly 239pU and
241PU. These isotopes are formed when the mU in the fuel cells capture a thermal neutron.
Historical Overview and Properties of Radioactive Waste 18
The main reason for this is that these isotopes are also fissionable and can therefore also be
used in the production of new fuel elements instead of the 235U.Indeed, they are often more
economical to use than mU, since their fission eftïciency is higher, and they differ chemi-
cally completely from uranium, so that they can be separated chemically from the uranium.
Unfortunately, this process generates waste containing considerable quantities of uranium
fission products and highly energetic actinides.
2.3.3 Decommissioning of Nuclear Facilities
All nuclear facilities, especially nuclear reactors have a finite operational life and need to be
decommissioned at some point in time (lAEA, 1992). Since the facilities and their sites are
often highly contaminated with radioactive materials, they must be dismantled and decon-
taminated, before the site can be used for other uses, thereby generating what is known as
decommissioned waste. In the case of a nuclear reactor the volume of this waste, which
consists mainly of the construction materials of the dismantled facility, its hardware and
contaminated soil, is often much larger than the volume of waste produced by the nuclear
fuel cycle. The decommissioning of these plants is consequently often delayed, because of
the lack of suitable disposal facilities.
2.3.4 Other Forms of Nuclear Waste
The volume of radioactive waste generated by the industry, research laboratories and the
medical profession is small compared to the nuclear fuel cycle, but still may pose a consid-
erable health risk to humans if not disposed properly. Of considerable importance in this
regard are spent sealed radiation sources, or sealed sources for short, defined by the Interna-
tional Standards Organisation (ISO) as IAEA (1991):
'A radioactive source sealed in a capsule or having a bonded cover, the capsule or
cover being strong enough to prevent contact with and dispersion of radioactive
material under normal conditions of use and wear for which it was designed.'
The applications of sealed sources found in the industry include: belt, density, level and
thickness gauges, industrial radiography, moisture detectors, the sterilisation and preserva-
tion of food, and roentgen t1uorescent analysis. Research applications include calibration
sources, electron capture detectors, densitometers, tritium targets and eliminators for static
electricity. Sealed sources are also used widely in the medical field for bone, brachytherapy,
teletherapy and clinical radiotherapy (lAEA, 1991). Since the physical dimensions of these
sources are small, although some of them may contain highly active isotopes, they are very
susceptible to theft and misuse, with a corresponding potential radiation hazard tohumans,
Historical Overview and Properties of Radioactive Waste 19
Magnitude
o(Problem
Causingmost 1 Brachytherapy 1 1 Industrial. 1 Tele·Radiography therapy
Problems
- - - - -
r Moisture I - - - - -
Detectors
Causing [ Well LOllingProblems 1
r Industrial Gauges -I
- - - - - - - - - - - - - - -
Normally no I .rradiators forProblems Research and Industry 1
- - - - - - - - - - - - - -
Little I Calibration SourcesConcern I
- - - I - - - - - - - - - - - -No I ConsumerProducts
Concern
I I I I I
I kBq I MBq I GBq ITBq I PBq Source Activity
Medium
Very Weak Sources I Weak Sources I ISources Strong Sources IVeSry Strong Iources
Figure 2-2 Activity range for some important applications of sealed sources and the
magnitude of problems associated with spent sources (after IAEA, 1991).
Figure 2-2 gives an overview of sealed sources, their main areas of application and the
relative magnitudes of the problems associated with the different types of sources.
2.4 CLASSIFICATION OF RADIOACTIVE WASTE
2.4.1 General
All wastes that contain radioactive materials should be regarded as radioactive from the
physical point of view. However, the activities of the radioactive materials in some of the
wastes are so low that they do not present radiological hazards to the environment. The
!AEA (l993a) therefore define radioactive waste for legal and regulatory purposes as:
'waste that contains radioactive materials with activities higher than the clearance
levels established by the regulatory body, and for which no further use is foreseen.'
The preceding discussion indicates that it will be wrong to treat the various forms of radio-
active wastes on the same footing. The physical, chemical, biological and radiological prop-
erties of radioactive waste have consequently been used to devise criteria, summarised in
Table 2-2, for the qualitative and quantitative classification of the waste. Since these classi-
fication schemes are very useful in the management of radioactive waste, they are discussed
in more detail below.
Historical Overview and Properties of Radioactive "Vaste 20
Table 2-2 Important properties of radioactive waste used as criteria to classify them
(IAEA, 1994a).
General Specific
Origin
Criticality
Radiological properties Half life
Heat generation
Intensity of penetrating radiation
Activity of the radionuclides
Surface contamination
Dose factors of the relevant radionuclides
Physical properties Physical state (solid, liquid or gaseous)
Size and weight
Compactability hazards
Dispensability
Volatility
Solubility, miscibility
Chemical properties Potential chemical hazard
Corrosion resistance/corrosiveness
Organic content
Combustibility
Reactivity
Gas generation
Sorption of radionuclides
Biological properties Potential biological hazards
2.4.2 Qualitative Classification
There are a number of ways in which radioactive waste can be classified qualitatively. One
approach is to group the waste in terms of its origin, physical state (solid, liquid or gas), or
activity. A scheme based on the last property is particularly useful, as it allows one to clas-
sify the waste semi-qualitatively into low-level (LLW), intermediate-level (ILW) and high-
level (HLW) waste. This activity classification system is consequently the one most widely
used today.
In the activity classification LLW is waste whose activity is so low that it does not require
shielding during its handling and transportation, while ILW is waste that requires shielding,
but will not generate a significant amount of heat. High-level waste is, as its name implies,
waste whose activity is so high that it can generate a significant amount of heat ~ 2 kW rrr").
The best known forms of high-level waste include: the highly radioactive liquids, contain-
ing fission products and some of the actinides, generated by the chemical reprocessing of
spent fuel, and spent fuel itself, if declared a waste.
The IAEA (1994a) tends to differentiate further between short-lived, long-lived and alpha
bearing low-level and intermediate-level wastes. As used here, the terms short-lived and
Historical Overview and Properties of Radioactive Waste 21
)~.~------------------------------------~
HIGH LEVEL WASTE (HLW)
Heat Generation> 2 kW mo)
LOW- AND INTERMEDIATE-LEVEL WASTE (LILW)
Heat Generation < 2 kW mo)
...... Surface Dose Rate> 2.0 mSv h-I
en ,......,
"C E
.J .... ..,...... ;
:::I
w ''._"". ,-., &\,> .
E~ « 2%) « 1%)
...I f
W '" v Qj E.J ~ >C! 101 ...... ïoo>- ;.J
Is- :...J...
~...ï..o. o E 00
rT :::I'" ...I
III
Qj ..,.... III E 0
i= ~'" 0 .~ 0~ 0 .0.., 0 LILW (I) - SL i LlLW (I) - LLCJ :::I - ;.J
< 0 v E 0..., !'" .:_J. .v_.Q -al ...s ...
ï
~ oo._
V.
Qj
C ._. '" ..Q.... 00j ~ LlLW (L) - SL LILW (L) - SL101 101 IIIQj Qj0
~ "' ~ 0 tC" "C 'Ë
Qj 0Qj Qj ..., 'Ë
~..... .i!:: ~i: "C C"l:l.Qj
~0 ~0 .i!:: al (85%) « 15%)
'" '" 1 ~c0 Do
0 C
...I 0
...I Surface Dose Rate> 2.0 mSv h-I
CLEARED WASTE (CW)
DECAY PERIODS
Figure 2-3 The qualitative radioactive waste classification scheme proposed for the AEC.
The figure in brackets represent the estimated percentage of AEC waste
(AEC, 1997a).
long-lived waste refer respectively to radioactive waste that will, or will not, decay to ac-
ceptable activity levels, within the time period during which administrative controls are
expected to last. Alpha bearing waste, on the other hand, is simply radioactive waste that
contains one or more a-emitting radionuclides, usually actinides, in quantities above the
acceptable limits established by the national regulatory body.
2.4.3 Quantitative Classification
In many cases, the classification of radioactive waste is related to the safety objectives set
by a regulatory body for their management. These safety objectives are often formulated in
numerical terms, such a dose rate or activity level, which requires a quantitative classifica-
tion system (IAEA, 1994a).
The quantitative radioactive wa te classification scheme at the AEC, illustrated chematically
Historical Overview and Properties of Radioactive Waste 22
in Figure 2-3, recognises three main categories of waste: cleared waste (CW), low- and
intermediate-level waste (LILW) and high-level waste (HLW) (AEC, 1997a). Cleared waste
contains so little radioactive material that it cannot be considered as radioactive, and might
be cleared from nuclear regulatory control. That is to say, although still radioactive from a
physical point of view, this waste may be safely disposed of, applying conventional tech-
niques and systems, without specifically considering its radioactive properties (IAEA, 1994a).
In the AEC classification LILW can contain a wide range of radionuclides and radionuclide
concentrations that may require special packaging, conditioning and disposal options. Low-
and intermediate-level waste are consequently further subdivided into the sub-classes indi-
cated in Figure 2-3, and summarised below.
(a) Short-lived waste, sealed sources [LILW-SL(SS)].
(b) Short-lived waste; low dose rate [LILW(L)-SL].
(c) Short-lived waste, intermediate dose rate [LILW(I)-SL].
(d) Long-lived waste; low dose rate [LILW(L)-LL].
(e) Long-lived waste, intermediate dose rate [LILW(I)-LL].
However, only the qualitative classification scheme radioactive waste will be used in the
discussions that follow.
CHAPTER3
MANAGEMENT OF RADIOACTIVE WASTE
3.1 INTRODUCTION
Radioactive waste were often stored in the early days for an interim period to allow for
some decay, and then disposed of by dispersion and dilution in natural reservoirs (!AEA,
1993b). However, the production of larger waste quantities, such as the by-products of the
nuclear fuel cycle, made it necessary to develop new management strategies. Which strat-
egy to follow will, however, depend largely on the properties of the waste and the interna-
tionally accepted fundamental principles for the management of radioactive waste, as out-
lined in Chapter 1. This chapter will consequently be devoted to an overview of an inte-
grated radioactive waste management approach in Section 3.2, followed by a discussion on
some of the more important strategies that have been advanced over the years for the man-
agement of radioactive wastes, in Section 3.3. This is followed by a discussion of the main
objectives for radiation protection in Section 3.4, and a brief overview of the present status
of disposal practices for low-level, intermediate-level and high-level wastes, as well as spent
sealed sources, in Section 3.5.
All facilities associated with nuclear activities eventually have to be decontaminated,
decommissioned and the radioactive waste produced by the activities managed in a way
that obey the fundamental principles of radioactive waste management, as discussed in
Chapter 1. The cost and responsibility for these operations create what is called nuclear
liabilities (OECD/NEA, 1996), which is discussed in Section 3.6.
3.2 AN INTEGRATED APPROACH TO RADIOACTIVE WASTE MANAGEMENT
3.2.1 General
Radioactive waste management is the concluding part of all activities related to the produc-
tion of nuclear fuel, the generation of nuclear power, research and development, and the
many applications of radioisotopes, as discussed in Section 2.3 (IAEA, 1992). According to
the !AEA glossary (IAEA, 1993a), the term integrated approach refers to
'a logical and preferably optimized strategy of a radioactive waste management
programme as a whole, from waste generation to disposal, so that the interac-
-23-
Management of Radioactive Waste 24
tions between the various stages of waste management are taken into account
and that decisions made at one stage do not foreclose certain alternatives at a
subsequent stage.'
This definition is a confirmation of the eighth principle in Chapter I-radioactive waste
generation and management interdependencies-stating that interdependencies among all
steps in radioactive waste generation and management shall be taken into account. In many
cases the waste generating process is chosen before planning how the waste will be man-
aged. It has been recognized, however, that a careful consideration of the waste generation
from the beginning can help considerably in solving problems that may develop afterwards
(IAEA, 1992).
As mentioned in Section 1.2, the pre-closure and post-closure assessments have to be treated
differently, but the periods themselves are interrelated. For example, in the pre-closure pe-
riod, safety of the operators has to be assured and their radiation exposure kept at a mini-
mum. After closure of the repository, the main concern is to minimise the exposure of the
population to radiation. However, the treatment process of waste can determine the suitabil-
ity of waste for storage and the selection of a disposal option; conversely, the characteristics
of a chosen site or repository can determine requirements on waste specifications. It there-
fore makes sense to consider the various stages of an integrated radioactive waste manage-
ment process as part of a structured methodology for a post-closure safety assessment. This
includes waste generation, waste pre-treatment, waste treatment and conditioning, interim
storage, and disposal. These steps, taken prior to disposal, are very important, since they
determine the quality and quantity of waste to be disposed and the method that can be used
for the disposal. They and the sources and types of radioactive waste generated in the nu-
clear industry are discussed in more detail below. Two other, equally important, compo-
nents in an integrated approach to radioactive waste management, that which will not be
discussed in this thesis, are the need for a national legal framework (IAEA, 1995c) and the
role of the regulatory body (Environmental Agency et al., 1997).
3.2.2 Waste Pre-treatment
Waste pre-treatment comprises all technical and administrative operations to collect, sepa-
rate, characterize, classify, package and document radioactive waste, according to the na-
tional waste classification procedures, taking the radiological content and characteristics of
the waste into account. The purpose of pre-treatment is to try to change the characteristics
of the waste in such a way that it can be handled more safely and economically in the future.
Itmay also include volume reduction and decontamination of contaminated materials. Note,
Management of Radioactive Waste 25
however, that this may lead to the production of secondary wastes, such as contaminated
filters, spent resins and sludge, which may have to be separated and pre-treated again.
3.2.3 Waste Treatment and Conditioning
Depending on the level of treatment done during the previous step, waste treatment encom-
passes the operations intended to benefit safety and/or economy by changing the character-
istics of the waste. This is done through volume reduction, removal of radionuclides from
the waste and change in composition in order to achieve the preferred approach of concen-
tration and containment, rather than dilution and dispersion in the environment.
After treatment, the waste is still not in a condition that it can be handled and transported or
disposed safely. To achieve this the waste is almost always immobilised first-a process
known as conditioning. The simplest way to achieve this is to mix the waste with a solidifiable
material, and place the mixture into a container that also serves as a mould during the
infilling and solidification of the waste mixture, or waste form. Solidifiable materials often
used for this purpose include synroc, glass or horosilicate for HLW, and cement, bitumen or
resins for LLW and ILW. The key parameters that need to be considered in selecting the
solidifiable material include: the distrihution stability of the radionuclides within the waste
form and the rate at which the material will decay physically and chemically. The contain-
ers used for this purpose, of which an example is shown in Figure 3-1, may vary from the
common 200 L steel drums to precisely engineered thick-walled containers, depending on
the nature and activities of the waste.
The use of waste containers has considerable advantages. They can facilitate the handling
of the waste form considerably, facilitate transport, provide radiation shielding during em-
placement operations and isolate the waste form from the ground water at the disposal site.
Since the last property is particularly important, waste containers are usually designed to
protect the waste from the ground water for a pre-defined period. It is therefore important
that properties such as the mechanical strength, resistance to corrosion, wall thickness, the
waste form and site characteristics be taken into account when designing such containers.
3.2.4 Organisational Responsibility
The responsibility for radioactive waste management can be given to a central organisation
or a group of organisations. ENRESA in Spain, for example, have responsibilities covering
the management of all the waste including the decommissioning of nuclear facilities (OECD/
NEA, 1996). These responsibilities include
Management of Radioactive Waste 26
Figure 3-1 Model of a typical container u ed for the immobilisation of low- and inter-
mediate-level waste (NAGRA, 1992).
(a) Handling and conditioning of radioactive waste;
(b) Siting, design, construction, operation and closing of interim and final disposal fa-
cilities for high, low and intermediate level waste;
(c) Management of activities derived from the decomrnissioning of nuclear and radio-
active installations;
(d) Setting-up of systems for collection, transfer and transportation of radioactive waste;
(e) Conditioning, when required, of the tailings arising from uranium mining and mill-
ing, in a safe and definitive manner;
(f) Ensuring the long-term management of any waste disposal facility;
(g) Carrying out the necessary technical, economical and financial studies taking into
account the deferred cost of radioactive waste management.
Integrated structures within one organisation like ENRESA will enhance the capability to
recognise interdependencies between the various steps and to avoid conflicting require-
Management of Radioactive Waste 27
ments that could compromise operational and long-term safety (IAEA, 1995a).
3.3 MANAGEMENT STRATEGIES
3.3.1 Fuel Recycling
The radioactive waste produced on earth is rarely of any economic value, except for the
spent fuel discharged from nuclear reactors. This fuel contains substantial quantities of
materials, notably plutonium and uranium that can be used again as sources of nuclear
energy. This process, generally referred to as fuel recycling, is illustrated schematically in
Figure 3-2. However, fuel recycling can only be achieved if one can extract the plutonium
and uranium from the spent fuel.
Enriched Production of Electricity
Uranium Fuel Using a Nuclear Reactor
.s.. Q)
0. ::J
V)LI..
Less Radioactive Solids Fuel Uranium and Fuel
Liquids and Glass Reprocessing Plutonium Fabrication
~ Highly Radioactive --,..
Liquid Waste
Immobilization Immobilization
(in cement or bitumen
for example) (in glass for example)
( Disposal J Disposal
Figure 3-2 The different stages in the recycling of spent fuel used.
Fuel reprocessing is a costly process (AECL, 1994a) and produces highly active liquid
waste containing fission and activation products, in addition to less active solid, liquid and
gaseous wastes. Since the highly active liquid waste is usually more hazardous than the
spent fuel itself, be it only for a limited time, special precautions have to be taken in handling
it. One method used for this purpose is to dissolve the highly active wa te in molten glass.
This solution is then cast into blocks, which are allowed to solidify-a process called
vitrification. The less active wa tes are likewise dissolved in cement or bitumen which is
then poured in metal drum to solidify (IAEA, 1993b).
Management of Radioactive Waste 28
Fuel recycling can be performed safely and practically on an economical basis in situations
where there is a demand for spent fuel. The method can therefore be used to manage spent
fuel to some extent, but it does not provide an effective solution to the problem of radioac-
tive waste in general.
3.3.2 Storage of Radioactive 'Vaste
Storage is another method often used in the management of radioactive waste. In this method
the waste is placed in an isolated and controlled facility with the intent to retrieve the waste
for exemption, reprocessing or disposal later (lAEA, 1993a). Storage is thus a temporary
measure that can be applied at any stage of the management process. Nevertheless, there are
sound technical and economical reasons why one would store radioactive waste. Spent fuel
is, for example, usually stored in water pools, because of the excellent cooling and shielding
properties of water. However, dry storage containers, in the form of concrete structures,
have been developed for the long-term storage of spent fuel in Canada, Germany and USA
(!AEA, 1992).
Another advantage of storage is that it can be used to reduce the quantity of volatile
.radionuclides in the waste, the strength of its radiation field and the production of decay
heat. Waste consisting of short-lived radionuclides is consequently often stored until the
activity has decayed to such an extent that the waste can be released within authorised limits
to the environment. However, storage must be often applied today, not for technical, but
socio-political reasons. This applies in particular to high-level waste, the disposal of which
is often vehemently opposed by the public.
Although storage is feasible and safe, it remains a temporary solution and does not replace
the need for the development and application of a long-term solution to the problem of
radioactive waste (IAEA, 1992).
3.3.3 Disposal of Radioactive Waste
The term disposal of radioactive waste is commonly used to denote the emplacement of the
waste in an approved, specified facility, also known as a repository, without the intention of
retrieving it (IAEA, 1993a). Extreme care must therefore be exercised in the design and
construction of such a repository and where it is placed, if one wants to satisfy the funda-
mental principles of radioactive waste management in Table 1-1.
Although there are other options, the disposal strategies in use today are usually limited to
Management of Radioactive Waste 29
near-surface disposal and geological disposal. Near-surface disposal is, per definition, the
disposal of waste on or below the earth's surface. The repositories used in near-surface
disposal systems usually consist either of an engineered disposal vault, situated below or
above the earth's surface, or a trench dug into the surface. Earth trenches are normally back-
filled with a suitable material and sealed off with an engineered, compacted clay cap of a
few metres thick, to prevent infiltrating rainwater from reaching the waste. These repositor-
ies are conventionally used for the disposal of low- and intermediate-level wastes.
A special case of near-surface disposal, that is becoming popular, is borehole disposal. In
this method the waste is emplaced in a borehole drilled to depths between 40 m to 100 m.
One example where this method is used, is the Greater Confinement Disposal (GCD) Facil-
ity at the Nevada Test Site in the USA, which is used for the disposal of high specific-
activity LLW (Price et al., 1996). Another example is the BOSS iliorehole disposal Of
,SpentSources) disposal concept that is currently under development in South Africa for the
disposal of spent sources (Van Blerk et al., 1999). These facilities are often considered as
near-surface facilities, but may be more appropriately called intermediate depth dis-
posal systems, since the boreholes are considerably deeper than most existing near-
surface facilities.
Geological disposal involves the isolation of the waste, through a system of engineered and
natural barriers, at depths up to several hundred metres, in geologically stable formations,
with low permeabilities. No such permanent repositories have yet been constructed, but the
experimental facilities usually consist of a vault excavated a few hundreds of metres deep in
the host rock, and divided into separate disposal chambers. Access to the chambers is achieved
through vertical shafts, extending from the soil surface to horizontal tunnels that connect
the disposal chambers with one another.
Although geological disposal can be used for the disposal of all radioactive waste, near-
surface and intermediate depth disposal are more economical in many cases. This is espe-
cially true in the case of the waste containing only short-lived radionuclides, or low levels of
long-lived radionuclides. Geological disposal is consequently mainly reserved for the dis-
posal of high-level waste and waste containing long-lived radionuclides.
Another disposal option that has been investigated extensively, but is not considered viable
today, is the disposal of radioactive waste in the sea. Here, one should distinguish between
disposal in the sea, the seabed and sub-seabed. Two interesting variants of this option that
have been proposed recently is to dispose the waste in the vicinity of subdueting plates, or
in stable mudtlats on the sea floor (Hollister and Nadis, 1998).
Management (If Radioactive Waste 30
It follows from the preceding discussion that much can be gained if site-specific character-
istics, such as the shielding power of the natural geological formations, are taken into ac-
count when designing a disposal facility. This power, unfortunately, varies considerably
from one formation to the next, with the result that it may be possible to dispose one type of
waste safely in one geological formation, but not in another. This means that one has to
design a set of so-called waste acceptance criteria for a repository, to ensure that the waste
will not pose unacceptable risks to future generations and the environment.
The introduction of waste acceptance criteria does not mean that society will not have to
implement a long-term institutional control of the repository, but that the risk posed by the
waste to humans and the natural environment will be limited if the controls fail. For exam-
ple, the possibility that radioactive wastes can escape from stable, low-permeable forma-
tions is small. Disposal is therefore, at least in principle, able to meet one of the main
objectives of radioactive waste management, in Chapter I-to minimize the burden of ra-
dioactive waste produced today on future generations. However, this does not imply that the
present generation, which derives a significant benefit from the use of radioactive materials,
can rely solely on the geological environment to protect future generations from the adverse
effects of radioactive waste.
The disposal of radioactive waste is, despite its advantages, a controversial subject on the
local, national and international level. It is therefore of the utmost importance to have a
legal framework and well established procedures to decide when and where the method can
be implemented. Such an approach has an additional advantage in that it can assist the
decision-makers in the evaluation of alternatives, and also ensure that all the relevant fac-
tors are addressed in the final decision. This approach will be particularly attractive in a
society able to assign a proper mandate and responsibilities to the authorities, and partakes
actively in the evaluation of the strategy and its implementation (Forsstrëm and Westerlind,
1997). Implementation of the international objectives for the disposal of radioactive waste,
that will now be discussed, can contribute significantly in developing such an approach.
3.4 RADIATION PROTECTION OBJECTIVES .
The aim of all protection objectives for the disposal of radioactive waste is to protect hu-
mans and the natural environment, now and in the future, from adverse effects of the wastes.
This implies that future generations should be afforded at least the same degree of radiation
protection as given to the public today and that the safety of wastes should not depend on
the active maintenance of the disposal system beyond a certain period.
Management of Radioactive Waste 31
One approach to achieve this aim is to have a clear and consistent set of radiation protection
goals to ensure that the risk to any member of the public, when exposed to the radiation
from radioactive wastes, will be as small as possible. The current international set of protec-
tion goals is embodied in the so-called 'Framework for Radiological Protection', which is based
on recommendations of the International Commission on Radiological Protection (ICRP, 1977).
The goals form the basis of the regulations for radiological protection implemented in many
countries and the safety standards of the IAEA (!AEA, 1989, OECD/NEA, 1977).
The basic components of the IeRP recommendations, as set out in Paragraph 112 of their
Publication 60, can be briefly summarized as follows (ICPR, 1997):
(a) Justification of a Practice - No practice involving exposures to radiation should be
adopted, unless it produces sufticient benefit to the exposed individual, or society, to
offset the radiation detriment it causes.
(b) Optimization of Protection - All reasonable steps should be taken to adjust the
protection for any particular source of radiation, so as to maximize its net benefit to
the society, subject to the prevailing economic and social factors.
(c) Application of the Individual Dose Limit - A limit should be applied to the dose
an individual may receive from exposures to radiation sources, other than medical
exposures.
To achieve this goal the ICRP (1997) recommended that the following procedures be fol-
lowed in the radiological protection of the public against waste disposal practices.
(a) The control of public exposure to waste disposal should be exercised by using con-
strained optimization of protection. To allow for exposures to multiple sources, the
maximum value of the constrained exposure to a single source should preferably be
less than 0.3 mSv a', while the total exposure should not exceed 1 mSv a-I.
(b) In situations where environmental monitoring is used to monitor the confinement of
radioactive wastes, derived restrictions should be developed for application to the
monitoring results. Since environmental monitoring is often used to assess the com-
bined implications of all the relevant practices, these restrictions should be based on
a dose to the critical group approaching 1mSv a-I.
The IAEA initiated in 1977 an integrated program for the geological disposal of radioactive
waste. One part of the program was to recommend basic requirements for the protection of
human health and the natural environment, and another to establish criteria for the under-
ground disposal of solid radioactive waste (lAEA, 1983). Two of the recommendations that
Management of Radioactive Waste 32
emerged from the program have interesting consequences for the disposal of high-level
waste (IAEA, 1989):
(a) Radiological Safety - Ensure the long-term radiological protection of humans and
the environment in accordance with the current internationally accepted radiation
protection principles.
(b) Responsibility to Future Generations - Isolate high-level waste from the environ-
ment over long time-scales, without imposing undue constraints on future genera-
tions, or require that they maintain the integrity of the disposal system.
The aim of the radiological safety objective is clearly not only to protect human health and
the natural environment from the potential harmful effects of the waste at the time a dis-
posal concept is implemented, but also far into the future (AECL, 1994a).
There are a number approaches that can be followed to satisfy the responsibility to future
generations (AECB, 1987). One is to restrict the disposal options to those that are techni-
cally and economically feasible, and socially acceptable, and do not rely on long-term insti-
tutional controls to ensure safety in the future.
Although not stated explicitly in these recommendations, the basic principles of the protec-
tion of the environment are clear (IAEA, 1992). The ecological balance of areas should not
be disturbed unnecessarily and the pressure on rare and endangered species should not be
increased. Known toxic materials should also not be released into the environment without
careful assessment of their potential for re-concentration or accumulation.
3.5 WASTE DISPOSAL PRACTICES
3.5.1 Low- and Intermediate Level Waste
Most low- and intermediate-level wastes around the world are disposed of in near-surface
disposal facilities, consisting of unlined trenches and pits or engineered concrete structures.
These facilities are usually not more than 20 m deep. Currently, however, there is a ten-
dency to dispose these wastes in deeper facilities. In the United Kingdom, for example, no
further shallow land repositories will be constructed, except for the expansion of the Drigg
facility (IAEA, 1993b).
Depending on factors such as the characteristics of the waste, the site, the sources, climatic
conditions and legislative requirements, different repository designs have been adopted in
Management of Radioactive Waste 33
different countries. In some countries, the conditioned waste is placed directly in contact
with the geological material, while others use special engineered structures, such as earth-
mounded concrete bunkers and concrete vaults below the surface (Phillip and Clifton, 1989).
Special caps, such as the one at the Intrusion Resistant Underground Structure facility (IRUS)
at Chalk River in Canada (Dolinar et al., 1996) are also used sometimes. Since the success
of a near-surface disposal facility will ultimately depend on the capability of the system to
prevent the migration of radionuclides, the precise structure of the repository may not be
that important. What is important is that the repository should minimize the contact time
between waste and percolating water or grounilwater. Systems with engineered barriers are
consequently becoming more common (IAEA, 1993b).
Deep disposal facilities for LLW and ILW have been or are being constructed and operated
in some countries. These facilities are still near-surface, although it makes more sense to
refer to them as intermediate depth disposal. In this case, disposal is performed in low
permeable geological formations, at depths of tens to hundreds of metres. Waste acceptance
criteria for these facilities can be quite different from near-surface disposal, since the desired
isolation capabilities of the system are much greater. Examples of such facilities include:
The Asses salt mine located in a salt diapir in Lower Saxony, Germany, where disposal
took place between 1967 and 1978 at a depth of 500 m for LLW and 700 m for !LW
(Brewitz, 1986).
The salt dome Morsleben facilities in the former German Democratic Republic (Ebel
and Richter, 1986).
The Forsmark facility in Sweden, which is excavated in crystalline rock underneath the
Baltic Sea with a rock cover of 60 m (Carlsson et al., 1990).
The Waste isolation Pilot Plant (WIPP) repository in the USA, constructed in a bedded
salt formation, about 660 m below the surface, for the disposal of transuranic waste
from defence activities (USDOE, 1990).
The Konrad iron mine in Germany, where the overlying formations consist mainly of
argillaceous sediments with a total thickness of several hundreds of metres. (Holtz,
1986).
The Olkiluoto repository in Finland, constructed in hard rock, at a depth of 50 to 100 m
(!AEA, 1993b).
Another example of intermediate depth disposal is the GCD Facility at the Nevada test Site
referred to above. In this facility the waste is emplaced in the bottom 15.2 m of a borehole, 3 m
in diameter and 36.6 m deep, while the rest is back-tilled with native alluvium (Price et al., 1996)
Management of Rudioactlve Waste 34
An alternative waste disposal concept is deep-sea disposal. This concept is technically
feasible with negligible environmental impact, in part because of the enormous dilution
potential of the sea. However, there is an international non-binding moratorium by the
London Dumping Convention on sea disposal of radioactive waste since 1983 (IAEA,
1993b).
3.5.2 High-LevelWaste
There is no real urgency to dispose HLW and spent fuel, since vitrified HLW and spent fuel
can be stored safely for many years (IAEA, 1993b). However, as mentioned in Section
3.3.2, storage is not a permanent solution. Various concepts for the disposal of HLW are
consequently being developed at the moment. Three concepts for the disposal of spent fuel
that have received some international attention recently (AECL, 1994a), include:
Transporting the waste into outer space.
Transmuting some or all of the radioactive nuclides in the waste to stable nuclides.
Geological disposal in stable, rarely disturbed geological mediums, such as glaciers,
deep-sea sediments, or rock below the seabed, and sediments or rock on land.
To transport an entire bundle of spent fuel (or reprocessed waste for that matter) into space
will be prohibitively expensive. Moreover, one cannot dismiss the probability of a launch
failure and the associated radiological risk it may pose for some parts of the earth.
Although transmutation is, strictly speaking, not a disposal method, it is included in this
discussion because it is the only method that has the potential to eliminate radionuclides
from the waste quickly. The transmutation technology available today is, unfortunately, not
very suitable for this purpose. Nevertheless, the method has considerable potential; particu-
larly if it can be applied in conjunction with the reprocessing of spent fuel.
The geological disposal of radioactive waste in glaciers and other ice sheets on earth is
feasible, but will probably not be acceptable to the public. Sea disposal of high-level waste
is also prohibited by the London Dumping Convention, despite the same attractive features,
such as the high dilution capacity of the sea (IAEA, 1993b). A related method, but not sea
disposal per se, is the one recently proposed by Hollister and Nadis (1998). The essence of
their proposal is to dispose the waste at depths ;:::4000 m in the geological stable and bio-
logical unproductive mudt1ats, poised in the middle of the larger tectonic plates on earth.
However, it may take time before the international community accepts this proposal, if ever.
Geological disposal on land is therefore the only remaining alternative.
Management of Radioactive Waste 35
No repository for the geological disposal of HLW or spent fuel is in operation yet. However,
many countries are investigating potential host rocks (IAEA, 1993b). Several countries
(Denmark, France, Germany, the Netherlands, Spain, the USA and the former USSR) are
considering salt rock, while others (Argentina, Canada, Finland, France, India, Japan, Spain,
Sweden, Switzerland, the UK and the USA) consider the more conventional igneous, meta-
morphic and sedimentary rocks. Countries such as Belgium and Italy also investigate
argillaceous sediments, with their low permeabilities, high sorption capacities (for most
radionuclides) and high plasticity.
Some countries have already developed underground research laboratories for this purpose
(Kickmaier and McKinley, 1996), while others have developed conceptual repositories
(AECL, 1994(/). One site that has been investigated in considerable detail is the Yucca
Mountain Site in Nevada, where the waste will be disposed in unsaturated tuff. The reposi-
tories for the envisaged sites will all be deep-seated and surrounded by multiple artificial
barriers, to ensure that any release of radioactivity to the environment will occur at an ac-
ceptably low rate.
3.5.3 Sperit Sealed Source Disposal
A concept for the disposal of the spent sealed sources discussed in Section 2.4.3 requires
special consideration for a number of reasons. Large numbers of spent sealed sources exist
in many countries, including the Russian Federation (former USSR) and developing coun-
tries. In some cases, no other nuclear activities exist in the country and therefore these
sources are the only radioactive waste that needs to be stored and disposed. Even countries
with nuclear fuel cycle facilities and with either operational or planned near-surface reposi-
tories find spent sources troublesome to manage, as they need to be handled separately from
other types of radioactive waste. For example, some of these sources have very high activi-
ties and cannot be disposed in near-surface facilities. However, the small volume of this
type of waste does not justify a geological disposal concept.
This situation prompted the AEC to introduce the BOSS disposal concept (Van Blerk et al.,
1999), which uses a borehole drilled down to a 100 m as repository and stainless steel
containers as waste packages. In this concept, the bottom 50 m is used as disposal zone,
while the remaining 50 m are back-tilled to prevent intrusion. This BOSS disposal concept,
which is still under development, will be discussed in more detail in Chapter 5. The initial
safety assessment of the concept indicates, nonetheless, that it is robust and provides a
viable solution for the disposal of spent sealed sources (Kozak, et al, 1999). The IAEA,
however, has not yet adopted an official position on the disposal of spent sources in boreholes.
Management of Radioactive Waste 36
3.6 NUCLEAR LIABILITIES
3.6.1 Definition of Term Nuclear Liabilities
The term nuclear liabilities is a fairly new concept that has not been applied very much in
the nuclear industry. There is therefore no universal consensus how it should be interpreted
and applied in the industry. The definition that will be used in this thesis is given in OECD!
NEA (1996) who defines nuclear liabilities as
" ... costs, which an organisation is expected to have to meet in the future as a
consequence of its current or past nuclear operations".
This definition contains carefully selected phrases with specific implications that need to be
interpreted very carefully. The term cost in the definition clearly refers to cost expected to
be incurred in future. This cost can he expressed in terms of either current money values
(CMV) or net present values (NPV). In the CMV approach future liabilities are evaluated in
terms of what it would cost to meet it today. This tigure is then adjusted annually for infla-
tion and periodically revised to take account of technological or regulatory changes affect-
ing the costs. The value ascribed to a future liability is clearly independent of the time at
which the expense will actually occur in this approach. In the NPV approach the current
cost of a liability is estimated and then projected onto the expected time frame. An estimate
is then made of the net present value of the cash needed at the time, based on an assumed
discount rate. The estimated costs in the NPV approach can of course also be revised to
account for inflation and technological change. However, timing is crucial in this method,
since the later a liability is incurred the more will it be discounted.
The role of the organisation in the definition will largely depend on the legal requirements
in a country. In principle, one can distinguish between generators of waste and operators of
waste. In South Africa the generator is the producer of waste and is also the legal owner of
the material, with the important implication that it should also bear the financial burden for
all consequences flowing from the ownership of the waste. The operator, on the other hand
is deemed to be the manager of the waste. His main responsibili ty is therefore to ensure that
the waste is safely handled on behalf of the waste generator.
The phrase, have to meet, stresses the obligation resting on the liable party to manage its
waste in accordance with legal requirements. The liable entity is also required to make
provision for the future management of these liabilities.
How the term future in a liability assessment framework is interpreted depends on certain
Management of Radioactive Waste 37
legal requirements and assumptions. For example, if the liability of a certain organisation is
considered to be discharged at the time the waste is disposed, then the State or an equivalent
institutional control body should accept any liabilities that may occur after disposal.
The term consequence in the definition simply acknowledge that any consequences of nu-
clear activities is a liability that needs to be accounted for.
Current or past nuclear operations suggest that one should distinguish between current
liabilities based on current activities, and historic liabilities based on past activities. Current
liabilities can be considered as liabilities created after a previous liability assessment. The
current activities should always be managed in an optimised way to ensure a minimised
current liability. The historic liabilities, incurred from past operations when inadequate pro-
visions were made for such obligations, can be reduced through carefully designed liability
reduction programmes. This can be enhanced, for example, through new waste manage-
ment techniques that may result in more effective waste management processes.
3.6.2 Sources of Nuclear Liabilities
The nuclear industry can be conventiently divided into five sectors: mining and milling, fuel
cycle activities, nuclear power plants, research centres and production of radioactive iso-
topes, for the purpose of identifying the various sources of nuclear liabilities.
Future liabilities in the mining and milling of uranium and thorium ores arise primarily from the
need to ultimately decommission the mine, mill, surface facilities and the waste management
areas. Although the volumes of waste in this sector are very large, the level of radioactivity is
rather low, with the result that the liabilities associated with the waste are relatively small.
The nuclear fuel cycle sector encompasses all nuclear fuel processing plants, including
facilities for the enrichment, conversion, production and reprocessing of fuel, as well as the
conditioning of spent fuel. Nuclear liabilities are incurred through the large volumes of
depleted uranium tailings, arising from the enrichment of uranium, and the decommissioning
of the fuel production plants. Liabilities also arise from the management of the produced
waste, such as reprocessing, vitrification, storage and disposal. Research centres and own-
ers of nuclear power plants, such as electricity utilities, face similar liabilities with the addi-
tion of liabilities associated with the management and disposal of the spent fuel, the low-
and intermediate level waste produced hy the plants and the decommissioning of the plants.
The nuclear liabilities of centres that produce high-level or long-life radioactive isotopes
Management of Radioactive Waste 38
(e.g., 60Co,99'fe,and 241Am),for the medical profession, research, industry and agriculture,
arise from waste produced during the production of the sources, the management and dis-
posal of spent sources and the decommissioning of the producing facilities.
It must be remembered though that liabilities may also arise after the waste disposal system
has been closed, or even abandoned. Since the activity of radioactive waste decreases with
time one can expect that the associated liabilities will also decrease with time, at a differen-
tial rate that increases slowly with time at first, but later more rapidly. Nonetheless, this
'post-closure' liability cannot be neglected and must be carefully considered in the post-
closure assessment of a radioactive waste disposal system.
3.6.3 Nuclear Liabilities Management
It is clear from the preceding discussion that the major sources of nuclear liabilities are the
costs of cleaning contaminated sites and facilities, and the costs associated with the man-
agement of the radioactive waste. These costs will clearly depend not only on present deci-
sions, but also on decisions taken earlier in the management of the radioactive waste. For
example, a site previously selected for the temporary storage of certain types of waste, may
eventually be used as a disposal site for all types of waste, which can place considerable
liabilities on the owner in future. Moreover, all activities performed within the safety as-
sessment framework will produce costs that must be accounted for in the calculation of
nuclear liabilities. It therefore makes sense to consider nuclear liabilities and the manage-
ment of nuclear liabilities as an integral part of radioactive waste management, if one does
not want to violate the eigth principle of radioactive waste disposal in Table 1-1.This ap-
plies in particular to the 'post-closure' liability introduced above. What one should guard
against, though, is to regard nuclear liabilities as more important than the fundamental ob-
ject of a post-closure assessment-the reasonable assurance that the disposed waste will
not affect future generations and the environment adversely.
CHAPTER4
COMPONENTS OF A RADIOACTIVE WASTE DISPOSAL SYSTEM
4.1 INTRODUCTION
The discussion in Chapter 3 has shown that disposal is the only method available for the
long-term management of radioactive waste. However, disposal as a selected strategy brings
about certain responsibilities and need to be dealt with within the framework set by the
principles of radioactive waste management. As shown by the discussion in Chapter 1, this
objective is best achieved by performing an integrated safety assessment of the disposal
system. As mentioned there, the discussion in this chapter will mainly concentrate on the
components of a disposal system and how they are incorporated in a safety assessment.
Thecomponents of a disposal system can be conveniently divided into two classes: internal
and external components. The internal components are those components that are situated
within the spatial and temporal boundaries of the system, while the external components are
situated outside these boundaries. These components can often be further divided into a
number of subsystems or components, that are linked to one another through various inter-
nal and external components or as it is more commonly known,features, events and proc-
esses (FEPs), as presented in Figure 4-1.
The external components in Figure 4-1 are usually the components over which people has
the least control. The discussion in Section 4.2 shows that these components are mainly the
result of natural and anthropogenic events induced on a regional scale.
The earth's surface has been used as a natural disposal site for anthropogenic wastes since
times immemorial, without any disastrous effects on the environment. One reason for this is
that the surface can easily absorb and decompose the waste, but only to a certain limit.
Unfortunately, this limit is often exceeded in situations where large volumes of hazardous
waste are confined to a small area, such as modern waste repositories. Some of the remain-
ing wastes may therefore dissolve in water that infiltrates the site to ultimately reach an
underlying aquifer, which may transport the dissolved wastes over vast distances and areas.
This migration of the dissolved wastes is fortunately a natural phenomenon that can be
restricted, to some extent, by applying one or more of the following procedures.
(a) Select a site with characteristics that will reduce the movement, for example where the
-39-
Components of a Radioactive Waste Disposal System 40
infiltration is small and the underlying geological formations are not very permeable.
(b) Study the physical, chemical, biological and radiological properties of the waste
that need to be disposed of in detail, and try to reduce its mobility by further processing
or conditioning.
Cc) Reduce the mobility of the waste with suitable engineered barriers.
A more detailed discussion of these procedures can be found in Section 4.3.
4.2 EXTERNAL COMPONENTS
4.2.1 Repository Factors
The repository factors in Figure 4-1 are those factors that need to be considered as part of
External Components
Repository Geological Climatic Future Human Other
Factors Processes and Processes and Actions and Factors
Events Events Behavior
Internal Componets
Site Waste Engineered Human
Characterization Characterization Barrier System Behaviour
Characterization Characterization
Biosphere
Dose to a Person
Figure 4-1 Conceptual representation of the different components of a disposal system,
and the flow of information between them.
Components (If a Radioactive Waste Disposal System 41
the planning, design, preparation, operation and final closure (e.g. capping) of the repository.
Some of the more important of these factors for a safety assessment are summarized below,
as a guideline for future investigators.
Cl The waste type(s) and volume(s) allocated to the repository.
e The design and physical dimensions of the repository.
• The scheduling and planning of events and activities that will occur during repository
excavation, construction, waste emplacement and closure;
• Quality control, assurance procedures and tests that have to be performed during the
design, construction and operation of the repository, and the form in which the waste
will be delivered.
• The contiguration in which the waste will be emplaced, and whether back filling
should be continuous or not in the case of earth trenches.
e The retention of records of the waste emplaced in the repository, and the placement
of markers at or near the site.
• Any special design, emplacement, operational or administrative measures needed to
enable (or ease) the retrieval of the waste, if such an option is applied or considered.
e The monitoring system needed to ensure that the repository will meet the design
criteria and that it will operate safely in the long-term.
o The cessation of the waste disposal operations at the repository and the subsequent
closure operations, including the capping and rehabilitation, if applicable.
4.2.2 Geological and Climatic Processes and Events
The time scales of importance to the safety assessment of radioactive waste disposal sys-
tems range from 1 year to 106 years. It is therefore important to consider what potential
influence geological and climatic changes, a few examples of which are given in Table 4-1,
may have on the long-term performance of a disposal system. Many of these processes
occur over geological time scales and may therefore not influence disposal systems for
short-lived radionuclides significantly.
4.2.3 Future Human Actions and Behaviour
One factor that is frequently ignored in the design of a disposal system, but may have a
major influence on its long-term performance, is the time span and degree of reliance that
can be placed on institutional control of the facility. Closely related to this factor are future
Components of a Radioactive Waste Disposal System 42
human actions and regional practices that can change the engineered barrier system and
geological barriers-thereby affecting the performance of the disposal system adversely.
These actions, of which a few examples is given in Table 4-2, may either be inadvertent or
deliberate and may occur during the institutional control period or thereafter.
4.2.4 Other Factors
There are a number of extra-terrestrial and other perceivable factors that can have a consid-
erable impact on the long-term performance of disposal systems. Two of these are comet
impacts and the development of new species that threaten the existence of mankind, espe-
cially in and around repositories. Although such events cannot be completely discarded, the
probability that they will occur is small. Moreover, the effects of some of the events, such as
comet impacts, will be far more devastating than just a disrupted radioactive disposal facil-
ity, and can only be combated by society as a whole. They will consequently not be consid-
ered further here.
4.3 INTERNAL COMPONENTS
.4.3.1 General
As mentioned above, the internal components are those components that are situated within
the spatial and temporal boundaries of the disposal system. According to Figure 4-1 these
components consist of the site information, physical, chemical, biological and radiological
properties of the waste, engineered barrier system information and human behaviour char-
acteristics that could influence the performance of the disposal system. These components
are used in combination with the external components to define the migration of dissolved
waste through what is known as the near-field, geosphere and biosphere. It is therefore
necessary to characterize these internal components to quantify the dose a future individual
may receive from the disposed waste.
There is no standard approach to the selection of a suitable site for radioactive waste dis-
posal. Seeking the best site is not always possible or mandatory. What is required is that the
site must meet the applicable national regulatory requirements. This means that the poten-
tial operator must demonstrate to the regulatory authority that the chosen site meets all the
necessary regulations, and in some countries that concepts offering obvious advantages
have not been overlooked. To achieve this the potential operator must obtain enough infor-
mation to evaluate the suitability and long-term performance of the proposed disposal facil-
ity (IAEA, 1993b). This process is referred to as site characterization in the literature and
Table 4-1 Examples of long-term geological and climatic factors that may influence the safety of a disposal site adversely.
Changes in the sea level, caused by geological and climatic changes Hydrothermal activities associated with high groundwate r temperatures
Ecological responses to a change in climate Long-term, global changes in the climate (from past, pre:sent and future evidence)
Extreme weather effects in tropical and desert areas Periglacial effects
Erosion and sedimentation on the geological scale Physical deformation of geological structures CoJ
3
Geomorphological changes Seismic events and their potential to occur ...,c
:~;,
Hydrological and geohydrological responses to climatic changes Tectonic movement and oregeny :;,
!7.
Hydrological and geohydrological responses to geological changes Volcanic and magmatic activities -:::.,
e;
::0
e;
c,
t$
Table 4-2 Examples of future human actions that can affect the post-closure safety assessment of a radioacti ve disposal facility adversely r..:..
-e
and therefore should be included in a regional site characterization. ~~e;
~~
Mining and other underground activities 9-
Changes in the social and institutional structures "S'"1
1ii
Non-intrusive site investigations, eg for human resources ~
IJJ
'<
Future drilling activities (in the repository or its surroundings) '~.."...
Surface activities (excavations, site development, and archaeology) 3
Future actions needed to remedy problems with the waste repository
Knowledge of the location of the repository (inadvertent or deliberate intrusion)
Humankind's influence on the climate (emission of greenhouse gasses, deforestation)
Groundwater and surface water management activities (groundwater withdrawal, dam construction, irrigation schernes)
-----
A
<.J
Components of a Radioactive Waste Disposal System 44
commences during the site selection phase.
4.3.2 Site Selection
Siting of radioactive waste disposal sites can be described as the identification of potential
disposal sites (site screening and selection) followed by detailed investigations at one or
more preferred sites (site characterization). Site screening is the identification of a number
of potential areas or sites that may be technically suitable and acceptable to the public for
disposal and that warrants the expenditure of resources for detailed investigations. The first
step in the selection of a site for the disposal of radioactive waste is to decide what disposal
method will be used. A site covered with unconsolidated sands will obviously not be suit-
able for near-surface disposal, but may be suitable for geological disposal, if underlain by
thick impermeable rocks at suitable depths. Site characterization provides the technical
information needed to design the disposal concept and to assess the performance of the
disposal system. One or more of these sites are then selected as the candidate site and
characterized in detail, with the view to convince the regulatory authority that the site is
suitable for a radioactive waste repository. Table 4-3 lists a few geological and surface
features that can be useful in the regional selection of a site. More detailed discussions on
the stages of site selection for geological and near-surface disposal facilities can be found in
IAEA (1994b) and IAEA (1994c) respectively.
Before leaving this discussion of site selection it must be pointed out that it will be difficult
to get a site selected without the active participation and co-operation ofthe local populations
and other interest groups in all four phases of the site selection. This objective, also referred
to as confidence building (see Section 6.6.7), may not be easy to achieve, particularly in the
Table 4-3 Examples of geological and surface factors that should be included in a re-
gional site characterization.
Geological resources of the area
Evolution of the geological structures in the area
Characteristics of the aqui fer systems in the area
Regional hydrology and geohydrology of the area
Characteristics of the regional weather and climate
Characteristics of the erosional and depositional evolution of the area
Properties and characteristics of large-scale geological discontinuities
Characteristics and evolution of the area's coastal and marine features
Chemical and geochemical characteristics of the regional geology and ground water
Characteristics of terrestrial water bodies (lakes, rivers and streams) and their evolution
Characteristics of terrestrial water bodies (lakes, rivers and streams) and their evolution
Components of a Radioactive Waste Disposal System 45
case of high-level wastes (Easterling and Kunreuther, 1995).
4.3.3 Site Characterization
Site characterization, usually starts with the selection of a specific site during site selection.
The level of activities during site characterization will therefore depend very much on the
work done during site selection. Site characterization may therefore in some cases be re-
stricted to a confirmation of the geohydrological conditions and geological sequence at the
chosen site. However, it is essential that a comprehensive characterization of at least one
site be performed, as this information is needed in the final assessment of the site (USDOE,
1988; IAEA, 1990).
Site characterization is usually performed in phases. The advantage of this approach is that
it allows one to evaluate the process continuously and to determine whether the site remains
suitable for further investigation. This not only enhances the inherent iterative nature of the
site's safety assessment, but also facilitates modifications to or redesigning the investiga-
tion programme.
The main purpose of a site characterization is to determine what effect the presence of a
repository will have on the environment and vice versa. It therefore mainly embodies the
study of the near-field, geosphere and biosphere, as indicated in Figure 4-1.
4.3.4 Waste Characterization
Radioactive waste was qualitatively classified as low-level, intermediate-level and high-
level waste in Section (2.4). However, the physical, chemical, biological and radiological
properties of the waste can vary considerably within each level. It is consequently essential
that a complete characterization of the waste be performed as part of a site assessment. Of
particular importance in this regard is to establish the need for the further adjustment, treat-
ment or conditioning of the waste, and its suitability for further handling, processing, stor-
age or disposal (lAEA, 1993a). Other factors that are also important include: the waste
type, its form (solid, liquid or gaseous), toxicity, level of radioactivity and radiation emitted
and the quantity oflong- and short-lived radionuclides in the waste (initial inventory) (IAEA,
1993b). A description of the initial inventory should include the decay chains, fission prod-
ucts and the half-lives of the daughter nuclides. In those cases where the reprocessing of the
waste is considered, the heat output and volumes of waste will also be important. The infor-
mation needed for the two radionuclides GOeoand 90Sr are presented as examples in
Table 4-4.
Table 4-4 Information needed for a safety assessment analysis of 90Sr and 60CO (Seitz et al., 1992).
Isotope 90Sr
Half-life 28.78 years
Primary radiation and energies fT,550keV o
Daughter Products Yttrium (90Y) Half-life: 2.671 d Emission: fT, 2.27 MeV o3
Major Chemical forms Sr++,s-eo, (s) ~:;
tt
Production methods Fission Product :::l
!t
Waste stream Nuclear power plants: Ion exchange resins, pre-coat filters, cartridge filters, concentrated liquid wastes, trash, dry .c...,
active waste and waste from decommissioning tol~
Major environmental pathway Milk, Groundwater tolc,
Health hazards Ingestion and internal exposure from beta emission ('bone seeker') c:;
-- -- - -- --- ---- ::l
-e
tt
Isotope 6OCo ~r.;-
:Jj
Half-life 5.271 3 years -tt
Primary radiation and energies gs:fT, 2.7 MeV; ml: IT 59 keV (lOO%),jr, 2.9 MeV (0.240%) !?:.n
'tl
Daughter Products Nickel (6°Ni) Stable o
~'"
Major Chemical forms Nr., NiOH , NiCI2, Ni(OHh NiS (s), NiCO) Cl)
'<
Production methods Produced by the neutron activation of 59CO ~
It>
Waste stream Nuclear power plants: ion exchange resins, concentrated liquids, filter sludge, cartridge filters, trash, dry active waste 3
Waste from medical and academic institutions, teletherapy units, radiography cameras and geological and
biological studies
Major environmental pathway Groundwater (highly retarded by sorptive processes)
Health hazards Direct exposure due to gamma emission, ingestion and internal exposure
"0"\"
Components of a Radioactive Waste Disposal System 47
4.3.5 The Engineered Barrier System
Engineered barriers, constructed from materials with favourable properties and predictable
nuclide retention capacity, usually provide the primary containment of the waste in a mod-
ern disposal system. The possibility therefore exists that most radionuclides will decay to
insignificant levels within the barriers, if the disposal concept and the construction of these
barriers are implemented correctly (Smith et al., 1994). However, nothing on earth is per-
fect. These barriers should therefore be constructed in such ways that they will still limit the
release of nuclides to the geosphere in case they fail. This can only be achieved if the de-
signer has sufficient information on the physical, chemical and biological properties of the
proposed barrier materials, the surrounding near-field, geosphere and biosphere.
The first strategy conventionally employed to prevent the migration of the nuclides from the
repository is to condition the waste, as discussed in Section 3.4. The ability to design a
'perfect' container for a specific disposal site will, however, be limited by economic and
construction factors and the geometry of the waste form. The containers used in practice are
usually cylindrical in form, mainly for two reasons: the cylinder has the largest volume to
. surface area ratio of all solid figures, except the sphere, and it is relatively easy to handle.
However, this means that there will always be gaps between the containers and the reposi-
tory wall, when the containers are put in the repository. This could expose large parts of the
container to interactions with its immediate surroundings, thereby reducing its effective-
ness. Repositories are consequently usually back-tilled with suitable materials that will
reduce the interaction of the containers with their immediate surroundings as much as pos-
sible. Considerable care must therefore be exercised in choosing the backfill material. Some
of the more important interactions to consider in choosing the back-fill material are:
(a) the mobility of the nuclides from the repository and the corrosivity of the containers,
(b) the structural stability of the repository,
Cc) heat flow in and around the repository, especially in the case of HLW,
Cc) intrusion of the repository by humans and other animals.
It will be very difficult if not impossible to completely prevent the intrusion of humans and
animals with the matelials available today, but the intluence of the other interactions can be
readily reduced. One way to achieve this is to choose matelials with low hydraulic conductivities
but high sorptive capacities, and at the same time reduce the corrosivity of the groundwater.
A number of materials with these properties have already been studied intensively both in
the laboratory and in field CAECL, 1994a). However, it may not always be possible to use
Components of a Radioactive Waste Disposal System 48
the most suitable material for a specific site, because very little may be known of the mate-
rial's long-term performance, or the material may not be available in sufficient quantities.
(Several million cubic metres of material may be needed to fill a HLW repository, for exam-
ple.) Bentonite, fullers earth, and various clays, zeolites and cements are considered as the
best back-fill materials at the moment (IAEA, 1993b).
Infiltrating rainwater poses one of the biggest threats to the integrity of a near-surface re-
pository. It has therefore become practice to place an engineered cap on top of such facili-
ties to prevent rainwater from entering the system. Depending on the size of the facility,
these caps are normally not more than a few meters thick, and constructed in such a way that
most of the rainwater is diverted to a drain. Clay compacted to a very high density and with
a low permeability is a very popular material to use in engineered caps, However, clays
often tend to fracture, especially during dry periods, although the fractures may close again
if the clay is wetted sufficiently. Nevertheless it will be difficult, if not impossible, to ensure
the integrity of the cap indefinitely. However, near-surface repositories are mainly used for
low and intermediate level wastes, which need only he shielded for a limited time from
infiltrating rainwater. It may therefore be possible to design a suitable cap for such a reposi-
tory by assessing the behaviour of caps, constructed from different clays, under different
infiltration rates during the design phase of a repository.
It is rather unlikely that humans, burrowing animals or plant roots will never intrude a near-
surface repository in the future. Future engineered caps for these facilities must therefore
not only be infiltration resistant, but also intrusion resistant. The Intrusion Resistance Un-
derground Structure (IRUS) at Chalk River in Canada is an exam pIe of such a facility (Dolinar
et al., 1996).
4.3.6 Human Behaviour
The fundamental principles of radioactive waste management in Table 1-1 have one com-
mon objective: to protect humans and their environment from adverse effects of the waste.
It should therefore not he a surprise to learn that this is indeed the main concern of the safety
assessment of any radioactive waste disposal site.
Although some influences that humans may have on the integrity of a waste disposal site,
have been discussed above, such a safety assessment should also include identifying the
group of people (known as the critical group), most likely to be affected by the site, and
their local habits. The factors listed in Table 4-5 may serve as a guideline as to the type
information that should be gathered in this connection. For the time scales of concern, it is
Components of a Radioactive Waste Disposal System 49
virtually impossible to define the actual critical group. For this purpose, hypothetical criti-
cal groups are defined, based on past experiences and the factors listed in Table 4-5.
Table 4-5 Examples of human characteristics that should be included for the critical
group in the assessment of a radioactive waste disposal site.
Time spent in the environment
Community and social characteristics
Normal activities of tile group and their use of materials
Dietary patterns of the group mld their variation with age
Leisure and other activities associated with the environment
Rural, urban, agricultural mld industrial use of land and water
Information about the dwellings or other structures of the target group
Anatomical and physiological characteristics mld their variability with age
Activities a-;sociated Witil the processing ,Uldpreparation of food and water
4.3.7 The Near-Field
The near-field can be defined as the excavated area of the disposal system, including the
repository, its engineered banier and the part of the surrounding geology whose character-
istics have been or can be altered by the repository or its contents (IAEA, 1993a). The near-
field may therefore be considered as the source of all changes the repository will induce on
the environment.
As shown in Figure 4-1, the waste characteristics and the engineered barrier system will
mainly determine the strength of the source. However, there is another factor, not men-
tioned explicitly in Figure 4-1 that will int1uence the strength of the source-the location
and nature of the repository. As discussed above, the repository can consist of either a vault,
a trench in the earth's surface, or a borehole, which may be situated completely or partially
in either the unsaturated or saturated zone of the earth's surface. A detailed study of the
geohydrology of a disposal site's saturated zone, unsaturated zone, or both, must therefore
form an integral part of the site's characterization.
4.3.8 The Geosphere
The term geosphere is generally used to denote the host medium that surrounds the near-
field. This includes the soil and rock formations that over- and underlie the repository and
their water content. The geosphere therefore forms a natural barrier between the biosphere
and the near-field. This is the main reason why near-surface and geological disposal sys-
tems are such attractive options for the management of radioactive waste. Of particular
Components of a Radioactive Waste Disposal System 50
importance in this regard is the ability of the geosphere to maintain conditions favourable
for long-term waste isolation hy
(a) Limiting the rate at which contaminants can migrate from the near-field to the bio-
sphere, and
(b) Protect the repository from natural disruptions and human intrusion.
Special attention should therefore be paid to the geosphere during site selection and charac-
terization. A partial list of factors that are especially important in this regard is given in
Table 4-6.
The ability of the geosphere to act as a natural barrier is determined by a num ber of factors,
of which the most important are its hydrological and geochernical properties, and the physi-
cal isolation and stability of the disposed waste.
Two of the most important parameters to include in the characterization of the geosphere
are the ground water t1uxes and flow paths. The piezometric levels in the aquifer(s), the
hydraulic conductivities, storativities and porosities of the different stratigraphic units and
their structure, essentially determine these quantities. Geological environments whose
geohydrological properties are particularly suitable for the disposal of radioactive wastes
include (IAEA, 1993b):
e rocks with a very low water content and permeability (e.g. evaporates),
o fractured rock of low intrinsic permeability, in which radio nuclide transport would
be controlled hy the fracture network (e.g. crystalline rock in low relief terrain),
Table 4-6 Partial list of factors that are important in characterizing the geosphere of a
disposal site.
Topographic relief of the site
Stability of the geological formations
Thermal environment of the geosphere
Biology and biochemistry of the geosphere
Local topography and morphology of the site
The presence or absence of mineral resources
Geohydrology and geochemistry of the geosphere
Animal and plant intrusions that may disrupt the repository
Gas sources and production and their effect on the geosphere
Properties and characteristics of discontinuities in the geosphere
The stratigraphy of the rock formations, their physical and chemical properties
Mechanical deformation of the geosphere caused hy the construction, excavation and presence of the repository
Components of a Radioactive Waste Disposal System 51
o rocks with a potentially very low content of mobile water and permeability (e.g.
argillaceous rocks).
IJ unsaturated rocks and deposits.
The ability of the geosphere to act as a banier between the repository and natural environ-
ment can be further enhanced by the geochemical properties of the geosphere. This applies
in particular to the ability of the geochemistry to slow or prevent the degradation of engi-
neered barriers, and to limit the rate of mobilisation and transport of radionuclides, when
they are released from the near-field. The most favourable geochemical conditions for this
purpose are sites with a redoxing environment, that are well buffered by minerals in the
rock formations, have low concentrations of complexing ligunds, and groundwater with a
high pH-value. It must be remembered though that both the waste type and the repository
design will influence the geochemistry at a particular waste disposal site. This is particu-
larly important when engineered structures are expected to he stable for long periods, or
interact dynamically with the geosphere,
The geosphere can only be considered an effective natural harrier if it can isolate the wastes
over the time scale for which the repository was designed. Waste disposal sites, particularly
the sites for radioactive wastes, should therefore not be sited near active volcanic centres or
major fault zones, Attention should also he given to the possible neo-teetonic deformation
and the associated erosion rates of the site.
4.3.9 The Biosphere
The term biosphere is conventionally used to describe that part of the environment inhab-
ited by all living organisms, no matter what their form or origin is. However, there is a
tendency to resttiet the term to those parts of the earth's atmosphere, hydrosphere and
lithosphere associated with the activities of mankind, in the case of radioactive waste dis-
posal (IAEA, 1993b).
The geosphere is obviously the domain mostly exposed to radionuc1ides that may leak from
a repository. It is therefore unlikely that the radionuclides will affect most living organisms,
especially people, before they reach the biosphere. The biosphere therefore represents the
domain where living organisms will be exposed to radionuclides that have escaped from the
near-field and geosphere. Since the major objective of radioactive waste disposal is to pro-
tect living organisms from the adverse effects of the waste, it is important to include the
biosphere, be it only in the restricted form defined above, in the analysis of a disposal
system.
l
Components of a Radioactive Waste Disposal System 52
The nuclides may follow numerous paths before one or other organism in the biosphere
absorbs them. Torre and Simon (1997) devised the simple scheme of compartments in
Figure 4-2 to describe this migration. Radionuclides are assumed to be instantaneously
distributed within the compartments, but can be transported by different processes from one
compartment to another. The behaviour of the people that may be potentially affected by the
nuclides is particularly important in this regard.
Figure 4-2 Composition of the biosphere after Torres and Simon (1997).
Biosphere Transport Modelling
Atmosphere Soil
Surface Water Sediment
Physical Compartments
There are situations where air is the major transport mechanism for radionuclides in the
biosphere, for example, the movement of radon from tailings dams. However, water or the
hydrological cycle in general, will be dominant transport mechanism at most waste disposal
sites (Freeze and Cherry, 1979). Previous experience has shown, however, that the major
contributions come from bodies of water, or receptors as Torres and Simon (1997) call
them, such as boreholes, wells, watercourses (rivers, brooks, and streams), lakes and the
oceans.
Boreholes and wells, although small in comparison with the other receptors, are particu-
larly important for the transport of radioactive nuclides in the biosphere. The main reason
for this is that they are the only receptors situated completely within the geosphere. The
chances that the concentration of radioactive nuclides in their water will be diluted (as is
almost always the case with the other receptors) are therefore zero, unless the aquifer on the
site is multi-layered and not all the layers are equally contaminated. The travel time of
Components of a Radioactive Waste Disposal System 53
contaminated groundwater from the repository to boreholes and wells is usually also much
shorter than to the other receptors. These receptors may therefore become contaminated
long before the other receptors.
Table 4-7 Summary of physical processes that may cause contamination of the terres-
trial food chain by nuclides that leaked from a radioactive waste repository.
[After Torres and Simon (1997).]
Uptake by plant roots
Interception and retention by vegetation
Retention by vegetables and soil surfaces
Depositions by either dry or wet processes, or both
Direct ingestion of surface soil by grazing animals
Transfer of the nuclides from ground water to the terrestrial system
Transfer of the nuclides from surface water by droplets of water (e.g. mist)
Transfer of the nuclides from surface water to sedirnents and the aquatic biota
Processes responsible for the translocation of the nuclides from the deposition site
Transfer of the nuclides from the soil, air, water and vegetation into the milk and meat of grazing animals
There are a number of pathways along which radionuclides can move from the receptors to
humans. For example, by eating fish and other seafood, swimming in contaminated water-
ways, lakes or the sea, and exposure to contaminated sediments on riverbanks, and the
beaches of lakes and the sea. However, the most common pathways, particularly in areas
where a disposal site is most likely to be situated, are through the domestic consumption of
water and the food chain. A few examples of physical processes that are particularly impor-
tant in the terrestrial food chain are summarized in Table 4-7.
CHAPTERS
CONCEPTS FOR THE DISPOSAL OF RADIOACTIVE WASTE
5.1 INTRODUCTION
In this chapter, three examples of disposal concepts for radioactive waste will be discussed
before presenting a broader perspective of the basic principles of a post-closure safety as-
sessment.
As mentioned in Chapter 1Vaalputs is the only one of the two sites, currently used for the
disposal of radioactive waste in South Africa that underwent a detailed site selection and
site characterization programme. The following discussion therefore begins with a descrip-
tion of the near-surface concept used for the disposal of low- and intermediate level waste at
Vaalputs, in Section 5.2. This is followed by a discussion of the disposal concepts used at
the other site, Thabana, in Section 5.3.
A permanent solution for the high-level waste generated by the Koeberg Nuclear Power
Plant has not received much attention in South Africa, In fact, no conceptual plan of the
disposal concept that will be used in South Africa has yet been formulated. The general idea
now though is that the waste will be disposed at Vaalputs using a geological disposal con-
cept. This is a very important problem for the present generation to solve, if we do not want to
violate some of the fundamental principles of radioactive waste disposal, discussed in Chapter 1.
The geological characteristics of Vaalputs are very similar to the site proposed for geologi-
cal disposal in Canada by Atomic Energy of Canada Ltd. (AECL). A geological disposal
concept, based on the Environmental Impact Statement (EIS) prepared by AECL for the
disposal of nuclear fuel waste in Canada (AECL, 1994a, 1994b), is therefore discussed in
Section 5.4.
A large number of spent sources from the medical and other technical professions exist in
many countries all over the world. Some of these countries, many of them from Africa, find
it difficult to manage and dispose this waste. The AEC of South Africa therefore introduced
the BOSS (Borehole disposal Of s.pent s.ources) disposal concept as part of an IAEA pro-
gramme to strengthen the waste management infrastructure in African countries (AFRA 1-
14) (Van Blerk, et al., 1999). This disposal concept is described in Section 5.5.
- 54-
Concepts for the Disposal of Radioactive Waste 55
5.2 THE VAALPUTS NATIONAL RADIOACTIVE WASTE DISPOSAL FACILITY
5.2.1 General
In South Africa, the State accepts responsibility for disposing of radioactive waste, while
the AEC was instituted as the agent of the state for this purpose by the Nuclear Safety Act of
1993. The National Radioactive Waste Disposal Facility at Vaalputs is almost exclusively
being used for the disposal of low- and intermediate level waste from the Koeberg NPP. As
discussed in Chapter 1, it went through a detailed screening, selection and characterization
process, with the result that vast amounts of data and information are available. The screen-
ing phase commenced in 1979 and the following parameters were examined in detail (Moore
and Hambleton-Jones, 1996).
Agriculture production Climate and rainfall
Corrosion of groundwater Ecologically sensitive areas
Industrial growth potential Mineral occurrences
Population density Proximity to international boundaries
Seismic hazards Surface and groundwater hydrology
Vaalputs was identified as the most suitable site for the disposal of radioactive waste in
1983. This led to a more detailed site characterization programme that lasted from 1983 to
1986, after which the site came into operation. Characteristics that were identified as being
particular favourable for the disposal of radioactive waste are: its arid and remote nature,
very low population density, annual rainfall of only 80 mm year", an unsaturated zone of
50 - 65 m, and its clayey soil properties. This led to the decision that near-surface earth
trenches in the overlaying clayey soils will be sufficient for the disposal of low- and inter-
mediate level waste and that concrete vaults are not necessary enhance the repository's
safety.
5.2.2 Site Characteristics
The farm Vaalputs is situated on the edge of the western side of the escarpment that divide
the inland and coastal plains in South Africa. As shown in Figure 5-1, the site is bordered by
rugged, mountainous terrain on the west and the nat plains of the Bushmanland plateau,
with an elevation of approximately 1 000 mamslon the east. The repository is situated on
this plateau.
Vaalputs is located in a transition zone between the winter and summer rainfall areas of
South Africa with winter rainfall somewhat dominant. The annual average rainfall is 80 mm,
Concepts for the Disposal of Radioactive Waste 56
Figure 5-1 Space map (Bands 4,5,3) of the Vaalputs area. The dune fields are shown in
light green, while the disposal site is shown in red.
while the potential evaporation is high with an annual average of 2 100 mm (Redding and
Hutson, 1983). The local geology ofVaalputs, shown in Figure 5-2, indicates that Vaalputs
is situated on the Precambrian Crystalline rocks of the Namaqualand Metamorphic Com-
plex. The crystalline rocks form the basement are covered by younger sedimentary rocks.
Metamorphism transformed the original sedimentary and volcanic rocks to granite-gneisses
and metavolcanics.
The basement rocks belongs to the pre-tectonic Garies Subgroup of the O'Kiep Group and
consists of light coloured biotite gneiss and quartzofeldspathic rocks (Brynard, 1988). The
Vaalputs granite-gneiss is fine to medium crystalline with a uniform pinkish colour, with
small amounts of garnet and biotite (Andersen, 1996). Large-scale tectonism folded, thrusted
and fractured the bedrock extensively. The basement rocks were also intruded by granites,
which caused extensive re-melting. The fractured basement rocks at Vaalputs are poten-
tially important for geological disposal.
The near-surface disposal trenches are situated in the Vaalputs Formation. This formation
consists of sediments - typical eeolian sand, calcretized sandy, gritty clay and red to greyish
fluvial gritty, sandy clay containing gravel and quartz pebbles - that accumulated in the
Vaalputs Basin. Calcite veins cut through the whole Vaalputs sequence and are deeper than
Concepts for the Disposal of Radioactive Waste 57
Formation Lithology Geological Process Geological Process
Kalahari
Gordonia Red sand 20-5 Acolian dunes
Calcrere and siterere
Ferruginiscd sandy gritty clay
Vaalputs Brown sandy gritty clay 35-20 Fluvial Tertiary to Recent
Grey sandy gritty clay with
interbedded pebble bands
Siliceous sandstone
38-35 Alluvial fun
Basement granitoids
1050 Tectonism Pre-Cambrian
O'Kiep Intrusion
Garies Metamorphic roek
Figure 5-2 Geological succession at Vaalputs. (Modified from Levin, 1988)
7 m (Levin, 1988). These veins resulted from infilled fractures. The Vaalputs Formation is
overlain by north-east trending red Kalahari sand dunes of the Gordonia Formation that
conform to the underlying geological structures.
5.2.3 Waste Characteristics
As mentioned above, Vaalputs is almost exclusively being used for the disposal of low- and
intermediate level waste from the Koeberg nuclear power plant. It consists essentially of
trash, equipment, filters, ion-exchange resins, evaporator concentrates and liquid chemical
waste. The inventory for the low- and intermediate level waste disposed of at Vaalputs as at
May 1998, is presented in Table 5-1.
The waste acceptance criteria for Vaalputs (AEC, 1997b) states that only solid waste will be
accepted, while solidified or immobilized waste must contain no free-standing liquid. Waste
containing transuranic nuclides may be accepted provided that the specific activity of such
waste is less than 370 Bq g' and that they are evenly distributed within the waste form. The
maximum permitted surface contact dose rate for any single waste package is 2 mSv h-I.
Physically, the mass of any individual package shall not exceed 6.3 ton, while the maximum
dimension of any waste package in any direction, shall be 2 m. No explosives or pyrophoric
materials shall be dispatched to the site.
5.2.4 Waste Packages
The only packages used for disposal at Vaalputs to date are the metal drums and concrete
Concepts for the Disposal of Radioactive Waste . 58
Table 5-1 The low- and intermediate level waste inventory at Vaalputs in May 1998.
Isotope Low Level Waste (CEl)) Intermediate Level Waste (CEq)
Received Decayed Received Decayed
IIOAg 2.66E-04 1.3lE-06 7.66E+02 4.46E-Ol
14C 1.l2E-02 l.12E-02 2.60E-Ol 2.60E-Ol
6OCO l.l9E+04 5.93E+03 6.33E+04 2.43E+04
134CS 3.91E-02 6.58E-03 2.46E+03 2.l3E+02
137Cs 3.56E+03 3.l5E+03 1.13E+04 9.53E+03
134Ce O.OOE+OO O.OOE+OO 8.94E+02 1.51E+OO
3H 3.56E+02 2.64E+02 4.07E+03 2.73E+03
1291 9.62E-08 9.62E-08 3.7lE+OO 3.7lE+OO
54Mn 1.47E-02 2.l0E-04 1.43E+03 4.lOE+OO
90Sr 2.33E-04 2.l0E-04 1.43E+03 4.l0E+OO
99Tc 4.49E-08 4.49E-08 1.34E-03 1.34E-03
Tru 7.68E-06 7.68E-06 4.l3E+OO 4.l3E+OO
containers shown in Figure 5-3. The concrete containers are all 1.3 m in height, with an
outside diameter of 1.4 m. Depending on the waste characteristics the wall thickness will
vary between 150 and 400 mm to ensure proper shielding against radiation. The metal drums
are standard 200 I drums, with a height of 875 mm and an outside diameter of 585 mm.
The low-level waste is compacted in the metal drum, while cement is used to solidify the
intermediate level waste in the concrete container. The concrete container is fitted with a
steel liner and a concrete cap. A gap between the steel liner and the concrete wall allows for
expansion of the core, while the cement immobilising the waste is setting. Separate trenches
are used for the concrete containers and the metal drums. This arrangement simplifies in-
ventory control, since most of the activity is con tined to the concrete containers. It also adds
to the stability of the intermediate level waste, as these containers tend to degrade and
corrode less rapidly than the metal containers.
5.2.5 Physical Dimensions of the Waste Disposal Trenches.
The near surface trenches used for the disposal of low and intermediate level waste at Vaalputs
are 100 m long, 7.7 m deep and 20 m wide at the bottom and with sides that slope upwards
at an angle of 800, see Figure 5-4. The trenches are filled up from the bottom with a drain-
age layer of 0.2 m gavel, waste containers (5.2 m in height), a compacted clay cap of 2.0 m,
and a top sand layer of 0.3 m. Screened clay is used as backfill between the waste packages.
The original idea was to fill each trench completely before it is sealed off. However, the
Concepts for the Disposal of Radioactive Waste 59
" 600
Figure 5-3 The waste packages used for the disposal of radioactive waste at Vaalputs.
volume of waste received at Vaalputs is considerably less than expected during the design
and construction of the trenches. No trench has consequently been completely filled since
the depository began to accept waste in 1986. Since the weather conditions at Vaalputs
corroded the containers in the open more rapidly than originally anticipated, the trenches
are now divided into sections by concrete walls, as shown in Figure 5-4, and each section
backfilled and covered with sand when full. The trench will finally be provided with a
compacted clay cap to serve as an engineered barrier to rainwater infiltration. The cap will
be 1.5 m thick at the edge and 2.0 m at the trench centre, with a slope of 5% on its top
surface. It will be compacted in layers approximately 150 mm thick to achieve optimal
compaction. The cap will be covered with a layer of the original sand, at least 300 mm thick.
The whole burial area will finally be restored to its original topographic contours.
5.3 THE THABANA RADIOACTIVE WASTE DISPOSAL FACILITY
5.3.1 General
As discussed in Chapter 1, very little is known about the criteria used to select Thabana as
a disposal facility. The only information available is a report by Niebuhr et al. (1968). In
Concepts for the Disposal of Radioactive Waste 60
Figure 5-4 Near-surface earth trenches u ed for the disposal of low- and intermediate
level waste at Vaalputs.
Concepts for the Disposal of Radioactive Waste 61
that report they motivate Thabana as a permanent storage area for active solid waste, given
the geological and the ion-exchange propelties of soils (Van der Westhuizen and Grobbelaar,
1967). Based on this information the types of waste that will be disposed are stated, as well
as the method of disposal that is envisaged.
Although several studies have been performed in the vicinity of Thabana since 1968, the
level of detail of the site and facilities are still very limiting compared to Vaalputs. This
could possibly be attributed to the intention that the waste will be retrieved some time in
future, i.e. Thabana was always considered as a storage facility (Van As and Basson, 1969).
However, the viability of retrieving the waste from some of the repositories is questionable
at this stage. Yet, the AEC has not obtained a license from the CNS to dispose radioactive
waste at Thabana.
Key issues with regard to the Thabaria facility is its present adequacy as a storage facil-
ity, the uncertainty regarding its total inventory and its future status, for example should
it be upgraded to an approved disposal facility or rehabilitated and closed. In the mean-
while, the assumption will be made that the waste will not be retrieved in the near
future.
5.3.2 Site Characteristics
As shown in Chapter 1, Thabaria is situated at Pelindaba near Pretoria, the Headquarters of
the AEC. The topography is rugged with a high relief and with slopes to the east and west.
The mountainous terrain strikes east west with incised valleys that runs north south. Thabana
itself is situated on a topographic high, with a potentiometric surface at 50 to 60 m below
the collar height.
The general geology of Pelindaba consists of alternating shale and slate with interbedded
ferrugineous quartzite layers of the Tirneball Hill formation of the Pretoria Group. The
slates are finely laminated with a heterogeneously developed cleavage (Courtnage et al,
1996). The quartzite layers have a variable grain size and are between a metre and two
thick. These sediments dip to the north at an angle of between 23° and 34°. The quartzite,
which is more resistant to weathering forms prominent lidges and broken boulders. Two
intrusive dykes-a dolerite dyke approximately 10 m wide and a syenite dyke 3 -5 m wide-
is present just south of Thabana.
The average annual rainfall at Pelindaba is 621 mm with a minimum of 375 mm recorded in
1991/92 and a maximum of 1 196 mm recorded in 1995/96.
Concepts for the Disposal of Radioactive Waste 62
5.3.3 Waste Characteristics
The original specification for waste at Thabana made provision for the permanent storage
of waste that contains low levels of a-,f3-, and y-rays that is sealed in plastic, or absorbed on
vermiculite, (in the case of organic material) and compressed in drums. Provision were also
made for solidified high-level waste (e.g. solidified concentrates from the evaporators),
solid high-level waste (e.g. spent sources, irradiated samples) and 'absorbed high-level waste'
(Niebuhr et al., 1969).
As mentioned above, an inventory of what has actually been disposed at Thabana is one of
the important uncertainties at this stage. The wastes include uranium (and its daughters)
contaminated equipment, scrap, miscellaneous metal items and drums, plutonium mixed
with cement and placed in metal drums, spent sources, medical waste, and baboon and pig
carcasses used for medical purposes. These putrefiable carcasses, which contained only
short-lived isotopes, such as 51Cr,22Na,57CO,59Fe,12519,5Nband I03Ru,were mixed with lime
and placed in drums.
5.3.4 Physical Description of The Repositories
Various kinds of repositories have been used for disposal at Thabana. The repositories used
for the disposal of radioactive waste consist mainly of simple near-surface earth trenches
and near-surface engineered boreholes. The level of information available on these reposi-
tories is very limiting compared to modern day facilities. It can only be assumed that at the
time of implementation the level of information was thought to be sufficient. The trenches,
with depths that vary between 3 and 8 m, contain a variety of wastes of which some were
placed in drums, while others were simply packed loosely directly into the trench. Six of the
trenches are filled and backfilled with excavated material and capped with imported clay.
Two other trenches were in the process of being filled, when people purposefully intruded
Trench 7 in search of a lost condenser containing 2 tonnes of depleted uranium in May
1996. No waste has since then been disposed at Thabana. As shown in Figure 5-5 the waste
disposed in Trench 7 consists mainly of drums and contaminated scrap metal, which were
then backfilled with soil.
The inventory of the trench is uncertain and the properties of the host medium, containment,
backfill and cap material are largely unknown. The long-term strategy and performance of
the trench is also unknown, as well as the rationale behind the physical design of the trench
as a disposal concept. This trench is a typical example of an older disposal facility and an
illustration of how radioactive waste should not be disposed.
Concepts for the Disposal of Radioactive Waste 63
Figure 5-5 An excavated portion of Trench 7 at Thabana, which reflect a typical past
practice near- urface disposal concept.
Another disposal system employed at Thabana is an engineered near-surface borehole facility,
presented in Figure 5-6 (a). The facility, used for the disposal of highly active 6OCOpencil
sources consists of a borehole approximately I m wide and 9.4 m deep. Three parallel
stainless steel pipes, with inside diameters of 75 mm, were positioned in the borehole
embedded in high-density concrete. There are two 30° bends in the pipes, approximately 4
m from the top to prevent a direct radiation pathway. The facility is capped with a concrete
apron incorporating a hinged metal lid provided with a padlock. The sources (with an activity
of 8.9 x 1013 Bq (in 1976) were conditioned in stainless steel containers, similar to the one
shown in Figure 5-6 (b).
5.4 CONCEPT FOR THE DISPOSAL OF CANADA'S NUCLEAR FUEL WASTE
5.4.1 General
The selection of a site for the disposal of any high-level waste is a controversial subject with
extensive financial implications. Although many countrie have selection procedures in place,
not one spent fuel element has been disposed in any facility in the world today. As men-
tioned before, some countries have developed underground research laboratories at poten-
tial sites, while others have developed 'engineered conceptual repositories'. These are re-
positories whose details, including their feasibility and potential environmental effects, have
been planned and discussed in some detail, but not to the extent that construction can begin.
One example of the latter facilities is the concept developed by Atomic Energy of Canada
r-----Cover
_,----- Lead plug
-Ol. ~Padlock
y oo:::
~(')
"0
fiS
_l _ 2 x 30° Bends ë'
- .
"".I..
~:r
S o
""f" ~ Vi'
cr' "o0
..... 1 Inside diameter 75 mm '"
- !., o...,
I»
C'"l.o'
~ High density concrete I.»(':.
)
;.;..
~
- , ~'~"
_I_--_-------------f-----------!-------
~-I(m)-~
(a) (b)
Figure 5-6 The engineered near-surface borehole facility (a) and stainless steel container used for the disposal of spent 6OCO sources at
Thabana (b). I ~
Concepts for the Disposal of Radioactive Waste 65
Ltd. (AECL) for the disposal of Canada's nuclear fuel waste (AECL, 1994a), the character-
istics of which will now be discussed in more detail.
5.4.2 Waste Characteristics
In 1992, 15% of the electricity generated in Canada was produced using CANDU (CANada
Deuterium Uranium) nuclear reactors, which are fuelled by uranium dioxide (U02)' Three
provincial electric utilities, Ontario Hydro, Hydro-Quebec and New Brunswick Power, own
these reactors and the spent fuel removed from them. A limited amount of spent fuel, from
prototype power reactors that have been shut down, is owned by AECL.
The U02 is formed into ceramic pellets that can withstand the high temperatures, pressures
and radiation fields in a nuclear reactor. These pellets are placed inside a tube, called the
fuel sheath, which is made of a zirconium alloy. Each end of the pellet-filled tube is sealed
with a welded zirconium alloy plug to produce a fuel element. Up to 37 of these tubes is
welded to zirconium alloy end plates to make a fuel bundle, which is about 100 mm in
diameter and 500 mm long. It has a total mass of about 24 kg, of which about 19 kg is
uranium. A schematic representation of the CANDU fuel bundle is presented in Figure 5-7.
The ranges of radionuclides and activities present in the fuel bundle is typical of nuclear
reactor waste.
lE:~~._ Fuel.B ... ket
Tube
E
E
Packed
Particulate
Figure 5-7 Schematic representation of the CANDU fuel bundle and container.
Concepts for the Disposal of Radioactive Waste 66
5.4.3 The Host Medium
A decision was taken in 1972 that geological disposal offers the best prospect for the dis-
posal of Canada's nuclear fuel waste. However, in 1974 it was decided to direct the research
mainly towards disposal in crystalline plutonic rock prevalent in the area of the Canadian
Shield in Ontario, shown in Figure 5-8. AECL proposed that the waste be disposed in a
vault, which would eventually be sealed, several hundred metres below the surface.
Plutonic rock is formed deep in the earth by crystallization of magma and/or by chemical altera-
tion. It is often referred to as crystalline rock or intrusive igneous rock, of which the most
common type is granite. This hard rock formation, which is similar to the basement rock of
Vaalputs, will serve to protect the waste form, containers, and vault seals from natural disrup-
tions and human intrusion. It should also be able to maintain conditions in the vault that will be
favourable for the long-term isolation of the waste. However, the host medium will have to retard
the movement of any radionuclides that will ultimately be released from the vault.
D Canadian Shield
I,' ,', I Thick Shale
\
I D Salt Basins
,I
I
I
/
I
{
\,
\
\
Figure 5-8 The plutonic rock in the Canadian Shield has characteristics considered to
be technically favourable for the disposal of spent fuel in Canada (taken
from AECL, I994b).
Concepts for the Disposal of Radioactive Waste 67
5.4.4 Requirements for the Disposal Concept
The requirements of the disposal concept proposed by AECL are based on the results of a
public involvement program. This program included consultation with special interest groups,
discussions with focus groups, conducting public opinion surveys, and commissioning lit-
erature reviews and case studies on social issues. These requirements are clearly in line with
the principles of radioactive waste management discussed in Chapter 1. That is, human
health and the natural environment must be protected from radioactive waste and that the
burden placed on future generations must be minimized, depending on the social and eco-
nomic factors. The building of public confidence that started with the drafting of the re-
quirements should continue in such a way that there is scope for public involvement during
all stages of concept implementation. Finally, the disposal concept must be appropriate for
Canada, that is, compatible with the geographical features and economic factors.
It was recognized from the beginning that these objectives could only be fulfilled if a suit-
able technology and expertise is developed and implemented, from the siting of the disposal
site, to the closure of the disposal facility. This technology should
(a) not rely on long-term institutional controls as a necessary safety feature-that is, the
disposal facility should be passively safe after closure.
(b) currently available or readily achievable.
(c) be adaptable to a wide range of physical conditions and social requirements and to
potential changes in criteria, guidelines, and standards.
(d) include the ability to monitor the site from time to time.
(e) include the ability to retrieve the waste-the provisions needed to retrieve the waste,
however, should not compromise the facility's passive safety.
5.4.5 Features of the Disposal Concept
A major feature of the proposed concept is that the spent nuclear fuel will be disposed in a
geological setting, with multiple baniers to protect humans and the natural environment
from both radioactive and chemically toxic contaminants in the waste. These barriers, shown
in Figure 5-9, are the container, waste form, buffer, backfill, other vault seals and the geosphere.
Institutional controls should therefore not be necessary to maintain safety in the long-term.
The expected waste form would be either spent CANDU fuel, or solidified HLW from
reprocessing, if the spent fuel were reprocessed in the future. The form of the spent CANDU
Concepts for the Disposal of Radioactive Waste 68
Figure 5-9 Schematic representation of the natural and engineered barriers that forms
part of the geological disposal concept for the disposal of Canada's spent
fuel (taken from AECL, 1994b).
fuel will reduce is solubility considerably and therefore contribute significantly to the retention
of radioactive and chemically toxic contaminants under the expected disposal conditions.
Nevertheless, it was decided that all waste, whether it is CANDU fuel in its current state, or
liquid radioactive waste from reprocessing, would be sealed in a container to facilitate its
handling and isolate it further from the surrounding environment.
The container, known as a packed-particulate container, is constructed from ASTM Grade 2
titanium (6.35 mm thick) to provide a design lifetime of 500 years, based on its corrosion
resistance. The container, shown in Figure 5-7, is an enclosed cylindrical vessel of all-
welded construction, with a maximum outside diameter of 645 mm and an overall height of
2246 mm. Inside the container, a basket holds 72 spent CANDU fuel bundles in 4 vertical
stacked arrays of 18 bundles. The basket is constructed of carbon steel tubes arranged in
concentric circles. Glass beads are compacted around the fuel bundles inside the container
to support the container wall and enable the container to withstand the external pressure that
would be imposed on it in a disposal vault. This will ensure that the waste is completely
isolated during the operation of the disposal facility and until there is a substantial decrease
in the activity and heat output of the waste. (The activity of the radioactive waste will be
more than 200 000 times less than when it came out of the reactor after 500 years.)
The waste containers will be emplaced in the disposal vault that will be excavated at a depth
between 500 m to I 000 m, below the surface, in the plutonic rock of the Canadian Shield.
It consists of a network of horizontal tunnels and disposal rooms excavated deep in the
rocks, with five vertical shafts extending from the surface to the tunnels, as shown in Figure
Concepts for the Disposal of Radioactive Waste 69
5-10. The shafts provide for ventilation and for transportation of container casks, materials,
equipment, and personnel. The greater the depth, the greater the minimum possible trans-
port distance from the disposal rooms to the surface, and lower the likelihood of any natural
disruption or inadvertent human intrusion. However, the in situ temperature, stresses and
cost of construction increase with depth. The vault itself consists of rooms or in boreholes
drilled from the rooms.
To help achieve the overall objective of limiting the release of contaminants from the dis-
posal vault, a number of different types of vault seals shown in Figure 5-11 would be needed.
Each container in the disposal room would be surrounded by a buffer material, which would
most likely contain clay. The buffer around the container has two purposes. The first is to
limit the rate of corrosion of the container and the dissolution of the waste form (should
groundwater seep into the container). The second is to retard the movement of any contami-
nants released from the waste form and the container. Each room would be sealed with
backfill and other vault seals such as bulkheads, plugs, and grout made of clay-based or
cement-based material. These seals would fill the space in the rooms, keep the buffers and
containers securely in place, and retard the movement of any contaminants released from
the container, waste form and the buffer. All tunnels shafts and exploration boreholes would
ultimately be sealed in such a way that the disposal facility would be passively safe; that is,
long-term safety would not depend on institutional controls.
5.5 THE BOSS CONCEPT FOR THE DISPOSAL OF SPENT SOURCES
5.5.1 General
As mentioned above, large numbers of spent sources from the medical and other technical
professions exist in many countries, even countries that do not possess facilities related to
the nuclear fuel cycle, that have to be disposed. This is particularly the case in Africa, South
America and some members of the Russian Federation. Since these sources need to be
handled separately from the other types of radioactive waste, mainly because of their activ-
ity to volume ratio, countries (even those with access to operational repositories) find it
difficult to manage and dispose this waste. This has led to the use of boreholes as disposal
units for these spent sources by some members of the Russian Federation (Van Blerk et al.,
1999) and in South Africa (See Section 5.3). However, the relatively shallow boreholes
used by these countries are not suitable for the disposal of isotopes with long half-lives,
such as 226Raand 241Am-a defect that the BOSS disposal concept try to eliminate. Al-
though the concept is still under development and evaluation, using the safety assessment
approach discussed in this document, an initial post-closure safety assessment has shown
Upcut
Ventilation
Shafts oo
n:I
t'!>
'C
t;
0'
."".I:r..
t'!>
!oii.
'C
~e:.
o....
~:::c
oQ.~·n.:.(...
t'!>
~
'."...t'!>
Figure 5-10 Cutaway view of the disposal vault for the geological disposal facility proposed for the disposal of Canada's spent fuel (AECL, 1994b). -..Jo
Concepts for the Disposal of Radioactive Waste 71
Concm. Plug 111ghlyCompKted
Benlonlt. BlOcks
Groot.
•Grouting BOlchol s
IGroulodFraetu.Zone
Ctay-a"scd ~81
Shaft
Cement-Based Scat Dlscreto Fractures
Grout
B e~1I1t
Coner te
Sh." Cop
Room
Exploratlon
Borohot
Sh~1t
Tunnel
Upper
Bc flit
Lower
Bt'cklllt
Hlgnty Comp8c1ed
B ntonllo Otocks Grout
Figure 5-11 Schematic representation of the various vault seals used to limit the release
of contaminants from the di posal vault of a geological disposal concept
(AECL, 1994b).
Concepts for the Disposal of Radioactive Waste 72
that the concept is robust and viable (Kozak, et al., 1999).
5.5.2 Requirements of The Disposal Concept
The BOSS disposal concept is intended to provide a solution specifically for the disposal of
spent sources and therefore must take into consideration the size and number of the sources
(i.e. the volume) that needs to be disposed. Several countries world-wide lack a nuclear
infrastructure that can control sophisticated disposal facilities. What they need is a disposal
concept that is technically feasible and economically viable to implement, taking the equip-
ment at their disposal and the their financial situation into consideration. Nevertheless, it is
important that the concept should comply with the safety objectives of radioactive waste
management as discussed in Chapter 1. Particular attention should therefore be paid to the
design of the repository and its operational requirements. For example, it would be advan-
tageous if the waste could be emplaced routinely over the life-span of the repository, the
need to actively maintain the site after closure could be minimized, and human intrusion
(purposefully or inadvertently) could be difficult.
5.5.3 'Vaste Characteristics
The main aim of the BOSS disposal concept is to provide a solution for the disposal of both
short- and long-life radionuclides, particularly isotopes such as 22GRaand 241Am. Table 5-2,
lists a typical inventory of spent sources from a country for which the BOSS disposal con-
cept is implied.
Tablé 5-2 A typical inventory of spent sources for which the BOSS disposal concept is
implied. (All dimensions refer to the height and diameter of a cylinder.)
Isotope No. of Activity per Total Activity Dimensions Application
Sources Source (GBq) (GBq) (mm)
192Ir 22 3.700E+03 8,140E+04 3x3 Gammagraphy
6OCO 2 3.700E+00 7,400E+00 3.2x3 Level Gauges
S7CO 4 1.850E-04 7,400E-04 - Medicine
mCs 11 2.775E+OO 3,053E+OI 6x4 Gamma densitemeters
5 5.550E+OI 2,775E+02 6x4 Well Logging
24IAm 560 I.850E-04 I ,036E-0 I 60x5 Smoke Detectors
9 5.550E-04 4,995E-03
226Ra 6 3.700E-03 2,220E-02 100x35 Calibration
3 1.lIOE-04 3,330E-04 70x50 Teaching
238PU 1 3.700E+OO 3,700E+OO 20x50 Static Electricity Removal
24IAmlBe 5 1.850E+OO 9,2S0E+OO 10xIO Humidity gauge
Concepts for the Disposal of Radioactive Waste 73
5.5.4 Features of The Disposal Concept
The conceptual repository for the BOSS disposal concept consists of a standard 165 mm
diameter borehole drilled to a depth of 100 m, as shown in Figure 5-12, with a 150 mm
casing. However, wider and deeper (or shallower) boreholes can be drilled if required, de-
pending on site-specific conditions and the sources to dispose. Van Blerk et al. (1999) dis-
cuss some guidelines for drilling the boreholes. The borehole and screen should be sealed
off with a cement plug at the bottom to ensure that the disposal volume remains dry during
the operational period. The disposal area can also be fenced in to limit access, and a tempo-
rary site office erected if necessary.
The disposal of the waste packages will be limited to the bottom 50 m of the borehole and
the rest backfilled with concrete to close the repository and prevent human intrusion. Ce-
ment sludge will be poured over the waste package after its emplacement as backfill, illus-
trated in Figure 5-13.
Van Blerk et al. (1999) argued that a casing constructed from radiation resistant PVC might
be the most suitable for the purpose, due to its small mass per unit length and potential
shielding power (especially to neutrons). They also discuss the suitability of various drilling
methods and their advantages and disadvantages under African conditions.
m
Stratigraphic;
Units
Figure 5-12 Schematic presentation of the repository area of the BOSS disposal concept.
Concepts for the Disposal of Radioactive Waste 74
I
Concrete backfill
Cement sludge __
Waste packages __
Casing driven through
wet cement plug
Figure 5-13 Graphical illustration of the in situ repository configuration for the BOSS
disposal concept, showing the position of the waste packages and the use of
different materials. (Not to scale.)
It follows from the preceding discussion that the disposal concept is, by definition, a near-
surface disposal concept.
5.5.5 Design of the Disposal Waste Packages
The nature and construction of the waste packages can play a very significant role in the
safety concept of the BOSS disposal concept, especially as an engineered barrier, but also to
handle and transport the packages.
The approach followed in developing the BOSS disposal concept was to set a reference
waste package design. There is little doubt that 226Rais the most demanding isotope to
provide for in a disposal concept for spent sources. A waste package suitable for 226Ra
should therefore also be suitable for the other isotopes of importance, such as 6OCo,mCs,
and wArn.
Concepts for the Disposal of Radioactive Waste 75
5.5.6 Reference Waste Packages Design for 226Ra
A large proportion of spent sources consists of radium needles. These needles are tubes of
platinum, platinum-iridium, or (more rarely) gold, which contain soluble salts of radium
(IAEA, 1996). Since the needles often tend to leak (Al-Mughrabi, 1998), the IAEA (1996)
recommended that they be encapsulated, using the techniques described by the IAEA (1996)
and Al-Mughrabi (1998). The capsules recommended for this purpose are tubes with an
outside diameter of 21.23 mm, length of 110 mm and wall thickness of 2.77 mm, con-
structed from Type 304 stainless steel (Al-Mughrabi, 1998). Each of the capsules will be
placed in a cementatory waste form within another Type 304 stainless steel container with
outside diameter 114.3 mm, length 250 mm and wall thickness of 3.04 mm, as illustrated in
Figure 5-14.
Once the waste package is closed and without any leaks, it is ready for disposal, as depicted
in Figure 5-15. This means that each waste package will be separated form its neighbour by
a cement layer 750 mm in the borehole. A single package and its backfill will therefore
occupy a length of 1 m in the boreholes, so that 50 waste packages can be disposed of in a
borehole with the standard dimensions proposed for the BOSS disposal concept. However,
the exact number will be determined by the inventory and the site characteristics, as derived
from the waste acceptance criteria for the disposal facility.
Air gap between needle and
Stainless Steel Capsule
226RaNeedle
Hook for Lifting
and Handling
Stainless Steel Lid -~~I
, Cementitious~:-H--- Waste Form
,
304 Stainless Steel
--)~
Container
Figure 5-14 Various stages in the preparation of waste packages proposed for the dis-
posal of spent sources in the BOSS disposal concept.
Concepts for the Disposal of Radioactive Waste 76
There are a number of reasons for proposing stainless steel as the container material and
cement as wa te form for the reference design in the BOSS disposal concept. Stainless steel
is, for example more resistant to corrosion than carbon steel and passivated by high-pH
conditions. The general corrosion and pitting corrosion rates are therefore expected to be
low, so that the containers do not need a protective coating. Crack corrosion cannot be
excluded, however, particularly in ground water with a high chloride content. Agg et al.
(1993) estimate that stainless steel will in general corrode at a rate of 0.3-1.0 urn year"
under the alkalie and anaerobic conditions of a repository (Kozak et al, 1999).
The cement on the other hand will act as a barrier, first between the container and aggres-
sive chemicals (primarily chloride) and then the capsule and chemicals, thereby reducing
the initiation of corrosion on the capsule. The cement will also provide a physical barrier
-.-_Q----}
375 m·.• •
-- -1-'-
304 Stainless Steel Container
1000 mm 250 m (Diameter = 150 mm,
~:-II--W,IITh;ckness = 3.05 mm)
:7Sfm
-·-----l·· :
375 rm
1000 mm 250lm
Borehole Casing
Figure 5-15 Schematic depiction of the ill situ placement of waste packages with the
BOSS di posal concept.
Concepts for the Disposal of Radioactive Waste 77
through which leached radium must pass before being released into the geosphere. At the
same time, it may also act as a chemical buffer, thereby limiting the release of radium
intrinsically (Kozak et al, 1999).
CHAPTER6
BASIC PRINCIPLES OF A POST-CLOSURE SAFETY ASSESSMENT
6.1 INTRODUCTION
A post-closure safety assessment was defined in Section 1.2 as an attempt to quantify the
exposure or risk posed by a radioactive waste disposal site to future generations of humanity
and their environment. Intrinsically, to do this one can say that the observations needed for the
safety assessment of a site should be carried out for the life span of the proposed disposal
facility. However, this is neither physically possible nor desirable. The only viable approach
to perform a complete radiological safety assessment is to try to obtain as much observa-
tional data as possible, on a limited time scale, and then simulate the future behaviour of the
disposal system through what is known as a model. Although there are other methods, most
models are developed on a computer, because of the complexity of the system.
The term model is, unfortunately, so often misused that it is not always clear what is meant
by it. This applies in particular to the literature on groundwater models (De Marsily et al.,
1992; Konikowand Bredehoeft, 1992). There is reason to believe that the same can be said of
the term model, as used in safety assessments. A more precise definition of the term model,
as used in this thesis, is therefore given in Section 6.2 to avoid any confusion before a broader
perspective of the basic principles of such an assessment is given.
The worldwide interest in the safety assessment of radioactive dis~osal systems has led to
numerous proposals how such an assessment should be implemented. The approach pro-
posed here, a flow chart of which is shown in Figure 6-1, is based on the approach of the Co-
ordinated Research Programme (CRP) of the IAEA on the 'Improvement of Safety Assess-
ment Methodologiesfor Near-Surface Radioactive Waste Disposal Facilities (ISAM) '(!AEA,
1997a). Although this programme focuses on near-surface disposal facilities, the principles
are also valid for geological disposal, as shown by the discussion of the basic framework of
the approach in Section 6.6. However, it may be useful to discuss a few properties of a post-
closure safety assessment first. The discussion therefore begins with the nature of a post-
closure assessment in Section 6.3. This is followed by a discussion of the purpose of such an
assessment in Section 6.4, and a few characteristics that are unique to a post-closure assess-
ment in Section 6.5.
-78-
Basic Principles of a Post-closure Safety Assessment 79
Context of the
Safety Assessment
Description of the
Disposal System
Describe and Justify
the Site's
Future Evolution
Formulate and Implement
Models for
Consequence Analysis
z
Perform the o
Compare with Consequence Analysis
Assessment Criteria Reject
Interpretation Disposal System
of Results
Assess Further
Information
Needs
Accept
Disposal System
Figure 6-1 Flow diagram of the ISAM approach for the post-closure safety assessment
of near-surface radioactive waste disposal systems.
6.2 MODELS IN SAFETY ASSESSMENTS
6.2.1 General
Botha (1994) ascribes the confusion in the interpretation of the word model in
groundwater literature, to a desire to differentiate between what is known in the exact
sciences as a theory, and the application of the theory to a real-world realization of the
phenomenon for which the theory was developed. The simplest approach to try to elimi-
nate the confusion of the word model, as used in the groundwater literature and safety
assessments, is probably to look at how scientists in the exact sciences, particularly
physics, perceive a theory. The theories of groundwater motion and mass transport that
play a very important role in radioactive safety assessments will be used as examples in
the discussion that follows. However, it must be remembered that they are not the only
theories used in such an assessment.
Basic Principles of a Post-closure Safety Assessment 80
6.2.2 Development of A Physical Theory
A physical theory is in essence nothing more than an attempt by scientists to understand and
control the natural environment, using abstract, philosophical reasoning to explain a natural
phenomenon, that is a phenomenon observable through the human senses. The first step in
the development of a theory for a physical phenomenon is then also usually taken when the
curiosity of someone is aroused by a particular observation of a phenomenon, who then tries
to explain the behaviour of the phenomenon.
The ancient Mesopotamian, Egyptian and Greek civilizations already recognize that a physi-
cal phenomenon can be related to three basic measurable quantities-space, time and matter.
However, it was left to the English scientist, Sir Isaac Newton, to show how a combination of
these measured quantities, and the ability of humankind to reason abstractly, can be used to
what Botha (1994) calls the conceptualization of a natural phenomenon. The result of this
conceptualization is what is commonly known as a theory in the exact sciences.
Many attempts have been made in the past to try to base a theory just on the ability of human-
kind to reason abstractly, especially in subjects (e.g., groundwater) where it is difficult to
make observations. Such an approach can be an interesting mathematical exercise, but is with-
out real physical substance. A theory always consists of two inseparable components: non-trivial
observations on the phenomenon, and the ability of humankind to reason abstractly.
It is quite common in developing a theory for a phenomenon, to come across quantities or
interactions that play a basic role in the theory, but that cannot be measured directly. In such
cases, the interaction is often related to more directly observable and measurable quantities,
henceforth referred to as observables, through a number of so-called relational or constitu-
tive parameters.
An important property of a theory, which is not generally recognized, is that a theory always
contains at least one theoretical principle, orpostulate, deduced from particular observational
facts. This postulate, often referred to as a law of nature, or simply law, can be rejected or
modified, but never be verified experimentally or observationally. A well-known example of
such a theoretical principlein the study of groundwater motion is Darcy's law (Botha, 1994)
q(x,t) = -KV<p(x,t) (6.1)
where the hydraulic conductivity tensor, K, relates the gradient of the piezometric head, f,
to the mass flux, q, per unit time over a unit area. Darcy (1856) derived this law from
observations on the flow of water through a sand column.
Basic Principles of a Post-closure Safety Assessment 81
Another example of a law that is particularly important in the assessment of a disposal
system is the law of mass conservation. This law first enunciated by the father of modem
chemistry, Antoine Laurent Lavoisier, holds that matter cannot be created nor destroyed.
Lavoisier derived the law from numerous experiments with gasses and other elements.
6.2.3 The Mathematical Model
Although a law is crucial to a theory, it is often not suitable for a detailed study of the
general behaviour of a phenomenon, because some of its observables cannot be measured
easily. Darcy's law in Equation (6.1), for example, relates three quantities: the discharge
rate q(x, t), the piezometric head gradient Vf(x, t) and the hydraulic conductivity K(x, t) at
a specific point (x, t) in space and time to one another. The law can therefore be used to
determine the value of anyone of the three quantities, by measuring the other two. How-
ever, it is not easy to measure the Darcy velocity in the field, with the result that it will be
difficult to use this law for the study of the piezometric pressure head, as a function of space
and time. What is commonly done in such circumstances, is to supplement the law with
additional mathematical information to develop a set of equations, known as the governing
equation of the observable.
The governing equation for groundwater motion is usually derived by applying the law of
mass conservation. This yields the so-called continuity equation (Bear, 1972)
DJp8(x,t)] +V·[pq(x,t)] = 0 (6.2)
where 8 denotes the volumetric moisture content, p the density of the medium and Dt the
partial derivative with .respect to time. Substitution of Equation (6.1) into this equation
yields
pSoDJ<p(x,t) =V-[pKV<p(x,t)]+ pf(x,t)
or if it is assumed that the density of the fluid remains constant
SODt[<p(x,t)] = V·[KV<p(x,t)]+ f(x,t) (6.3)
where So is the storativity of the medium and fix, t) the strength of any sources or sinks that
may contribute to the flow. This differential equation can be solved for ~x, t), once the three
constitutive parameters So' K and fix, t) has been determined. It is in this sense that Equation
(6.3) is regarded as the governing equation of density indepedent groundwater motion.
The governing equation for density independent mass transport in ground water can be de-
Basic Principles of a Post-closure Safety Assessment 82
rived using similar lines of reasoning. Application of the law of mass conservation, Fick's
law for molecular diffusion and Darcy's law yield in this case the hydrodynamic dispersion
equation (Botha, 1996)
[BOIc+ PbDls] + q'V c = V'[BDV c] - .:t(Bc+ Pbs) + (co - c)f(x,t) (6.4)
where c is the volumetric mass concentration of the solute, s the mass fraction of dissolved
solids absorbed by the matrix, rb the dry bulk density of the matrix (including the adsorbed
solids), I the radioactive decay constant, D the dispersion coefficient, and Co the initial con-
centration of sources or sinks with strengthflx, t), as in Equation (6.3).
The dispersion coefficient, which is a second rank tensor similar to K, is usually expressed
in the form
D=Dh+D (6.5)
In
where Dm is known as the molecular diffusion tensor and Dhas the coefficient of mechanical
dispersion, or hydrodynamic dispersion tensor (Botha, 1996). The ij-th component of this
coefficient, which arises from the interaction between the dissolved solids and water, can be
expressed as (Bear, 1979)
v.v.
D~ = arv(x,t)D ..+ (aL - aT)-'_J-
IJ IJ v(x,t) (6.6)
where aL and aT are two parameters, known as the longitudinal and transverse dispersivities,
with dimensions [L], dI..Jthe Kronecker delta
s, 1 (i=j), = { 0 o1 '# ).)
and vj the ï-th component of the seepage velocity, vex, t), defined by the equation
v(x,t) = q(x,t)
B(x,t)
where q(x,t) and B(x,t) are defined in Equations (6.1) and (6.2).
As shown by the examples above, governing equations arise from a combination of the laws
that govern the behaviour of a phenomenon. Nowhere in the derivations of the governing
equations were any particular realizations of ground water flow or contamination used. The
governing equation of a physical phenomenon therefore does not depend on a specific reali-
zation of the phenomenon. It therefore seems logical to refer to the governing equation as
the mathematical model for the phenomenon (Botha, 1996).
Basic Principles of a Post-closure Safety Assessment 83
6.2.4 The Conceptual Model
The governing equations of most physical phenomenon are expressed in terms of partial
differential equations, such as Equations (6.3) and (6.4). As is well-known, a partial differ-
ential equation can only be solved if all the constitutive parameters are known over the full
domain for which the differential equation is defined and appropriate boundary and initial
conditions are prescribed along the boundary, and within the domain, respectively. To illus-
trate these concepts more concretely, consider the simple two-dimensional parabolic partial
differential equation
D(u(x,y,t) = D~u(x,y,t) +D~u(x,y,t) (6.7)
defined over the rectangular domain, n, shown in Figure 6-2. The formal integration of this
equation will yield a solution in the form of an equation with five unknown parameters, two
from each of the double integration over x and y respectively, and one from the integration
over t. In practical applications these parameters are usually determined by prescribing
values of £l(x, t) on the four sides of the boundary an of n, in Figure 6-2, hence the name
boundary conditions. The fifth parameter, which arises from the integration over time, can
be similarly determined if one equates the solution with a known value of u(x, y, t) at a time
say t = to' hence the name initial condition.
Domain ern
Boundary, an, of n
Figure 6-2 Schematic representation of the domain, n, and its boundary, an, for the
parabolic partial differential equation in Equation (6.7).
Boundary and initial conditions will clearly vary from one domain to another and can only
be derived once a domain is prescribed for the governing equation. The same applies to the
constitutive parameters-the hydraulic conductivity and storativity for an aquifer in frac-
tured granite will clearly differ from that of a coastal aquifer in unconsolidated sands. Since
Basic Principles of a Post-closure Safety Assessment 84
it is not always possible to determine the boundaries of a phenomenon and its associated
constitutive parameters uniquely in the environmental sciences, different investigators may
assign different values to these parameters, as is often the case in practise. This observation
led Botha (1996) to introduce the term conceptual model to describe the governing equa-
tion, or set of equations, with its associated constitutive parameters, boundary and initial (if
required) conditions, for a specific realization of a phenomenon.
6.2.5 Analytical and Numerical Models
A conceptual model is not of much importance in practical applications, unless one can
solve its governing equation explicitly, subject to the prescribed boundary and initial condi-
tions. Although a large number of the governing equations of practical interest can be solved
analytically, others can only be solved numerically. This led Botha (1994) to introduce the
terms analytical model and numerical model. In his original formulation Botha uses the
term analytical model to describe the solution of the governing equation in a conceptual
model, when obtained with analytical methods, and numerical model, when the solution is
obtained with numerical methods. However, recent developments in algebraic computer
codes, such as Mathematica (Wolfram, 1996), blur this distinction somewhat.
Analytical models have one particular attractive property-they are always exact. Unfortu-
nately, there are not many differential equations for which analytical solutions are known,
particularly for phenomena with heterogeneous constitutive parameters and irregular bounda-
ries. Moreover, many of the analytical models are expressed in terms of complex math-
ematical functions, which are difficult to compute. Since numerical models can handle such
situations with ease, they soon replaced analytical models in many branches of science and
engineering after the introduction of the electronic computer in 1945.
The advantage of numerical models, however, comes at a price in that they are not exact. Great
care must therefore be exercised to erisure that the models are accurate. Numerical models must
consequently often be restricted in space and time, unless one has access to large-scale comput-
ing facilities, which is not the case with analytical models. For example, Botha (1998) uses an
analytical model to simulate the transport of radionuclides in the geosphere of tailings dams for
periods up to 60 million years without any difficulties. Although this period is far beyond that
required by the licensing authority, the simulation gave a considerably better insight into mass
transport through the geospheres of the dams, than could have been derived from the available
data, or a numerical model based on the data. It may therefore be worthwhile to give more
attention to analytical models in the safety assessment of radioactive waste disposal systems, or
even ordinary waste disposal systems, than is done at present.
Basic Principles of a Post-closure Safety Assessment 85
6.3 THE NATUREOFA POST-CLOSURE SAFETYASSESSMENT
A post-closure safety assessment is often viewed as a scientific exercise, carried out by
scientists sitting in front of a computer, trying to predict the future, mainly because safety
assessments are concerned with the long-term performance of the disposal system. Others
see a post-closure safety assessment as a scientific exercise in which every physical aspect
of the disposal system need to be analysed and included in the analysis to predict the actual
behaviour of the disposal system over thousands of years. Another view is that it is simply
a scientific exercise to demonstrate that a site meets regulatory safety objectives. These
different opinions suggests that it may be worthwhile to study the carefully selected phrases
in the definition of a safety assessment in Section 1.2 in greater detail, since they have
specific implications for safety assessments.
The term iterative process in the definition simply means that one must expect that a safety
assessment will have to be repeated two or more consecutive times. The advantage of such
an approach is that it allows one to use information from the previous assessment to refine
the design of the system and the collection of additional data. It also reduces the tendency
that the assessment will focus on one component at the expense of others.
The term site-specific prospective evaluations emphasize the fact that the assessment should
include data from the actual disposal system assessed. Not with the intent to predict its
actual behaviour in the future, but rather to understand the behaviour of the system better
and to reflect the importance of specific components with respect to the compliance criteria.
The safety assessment of a radioactive disposal system is not an exact procedure. The term
reasonable assurance in the definition of a safety assessment emphasizes this inexact na-
ture of the procedure. What one really wants to achieve in such an assessment is to reach
defensible decisions on the extent to which the disposal system may comply with the regu-
latory criteria. A safety assessment is therefore more a decision tool to determine the condi-
tions for which reasonable assurance of compliance with safety objectives can be provided
than a method to predict the actual behaviour of a disposal system into the future. The
results will therefore be largely a function of the data, design and assumptions used in the
analysis. Changes in anyone of these conditions can change the conclusions of the assess-
ment (Kennedy, 1997).
6.4 PURPOSE OF A POST-CLOSURE SAFETYASSESSMENT
The discussion in the previous section raises the question: What is the purpose of a safety
Basic Principles of a Post-closure Safety Assessment 86
assessment if it cannot supply exact answers? The best way to answer this question is to
again look at the main objective of the management of radioactive waste in Chapter 1
'... to deal with radioactive waste in a manner that protects human health and the
environment now and in the future without imposing undue burdens on future gen-
erations.'
It is important to note that this objective does not say to shield the present and future genera-
tions completely from the effects of radioactive radiation, but to protect them from the ef-
fects of the radiation without imposing undue burdens on them. To achieve this, the IAEA
(1989) introduced three other principles, closely related to the fundamental principles for
the management of radioactive in Table 1-1. The first of these principles-effects in the
future- states that:
'The degree of waste isolation provided by a disposal facility shall be such that there
are no predictable future risks to human health or effects on the environment that
would be unacceptable today.'
A simple answer to the question above would thus simply be to quote this principle. How-
ever this raises another question: how does one know that the risks acceptable today will not
be detrimental to generations in the far future? On the other hand, the possibility also exists
that future generations may not be as susceptible to radioactive radiation as the present
generation. Radioactivity is a natural phenomenon to which all people on earth are exposed
daily and will be exposed to in the future. It is consequently simply impossible to shield
future generations completely from the effects of radioactive radiation. The real problem
today is therefore not radioactivity per se, but rather the concentration of radionuclides in
hazardous sources by the present generation. The application of the previous principle will
therefore not discriminate against future generations. Indeed, the applications of radioactiv-
ity is so beneficial to people that future generations may again turn to it, even if all radioac-
tive wastes are removed from the earth today.
The main purpose of a disposal system is to protect future generations and the environment
they are living in, at times, which cannot be controlled today. It is therefore necessary to
determine how a disposal facility, which was designed, constructed and implemented today,
will affect future generations. The last two of the three IAEA principles referred to above
addresses this point.
The first of these principles-independence of safety from institutional control-says:
"The safety of a disposal facility in the post-closure period shall not rely on active
Basic Principles of a Post-closure Safety Assessment 87
monitoring, surveillance, or other institutional controls or remedial actions after
the time when the control of the repository is relinquished."
This principle essentially means that it is the responsibility of the present generation to
develop the necessary technology and infrastructure to ensure that future generations will
not be exposed to a health risk if the site is not controlled institutionally. The site should
therefore be safe enough to ensure that people will not be exposed to a health risk, even if
the site is not monitored, or future generations are not aware of its existence.
The last of the three principles-burden on future generations
"The burden onfuture generations shall be minimized by safely disposing of radio-
active wastes at an appropriate time, technical, social and economic factors being
taken into account"
is based on the ethical consideration that the generation who receives the benefits of a
practice should also bear the responsibility to manage the resulting waste (IAEA, 1995a).
In simple terms, it means that today's generation should clean up there own waste and not
burdens future generations with it.
The implementation of a disposal system, unfortunately, does not depend only on the scien-
tific and technical knowledge, but also on economic and social factors. One sociological
factor-the unwillingness of people to have a radioactive disposal site in my backyard
(Easterling and Kunreuther, 1995)-is, for example, responsible for the fact that the devel-
opment of new disposal facilities has been postpone world-wide. It may therefore be very
difficult for the present generation to fulfil the previous principle.
The application of the previous principles forces a constraint on the management of radioactive
waste not imposed on that of other hazardous waste, although some of these wastes may pose a
similar (perhaps even greater) risk to future generations. The management of radioactive waste
therefore differs completely from any other scientific or technical activity (Kozak, 1998).
6.5 CHARACTERISTICS OF THE POST-CLOSURE SAFETYASSESSMENT
It follows from the previous discussion that the post-closure safety assessment of a radioac-
tive disposal system will have some unique characteristics, which will be briefly discussed,
before describing the methodology itself.
As defined in Section 1.2, a post-closure safety assessment is concerned with the distant
future, and not the present or even near present time. It may therefore have to cover a con-
Basic Principles of a Post-closure Safety Assessment 88
siderable period, as illustrated by the fact that high-level waste disposed today may affect
the next 700 generations (Easterling and Kunreuther, 1995). Since the demographic fea-
tures of future generations and the environment are unknown, a post-closure safety assess-
ment can be interpreted as being more concerned with the protection of hypothetical per-
sons and a hypothetical environment than real people and a real environment. There is
therefore no guarantee that the principles valid today, such as the radiological protection
standards of the ICRP, will still be valid at the times covered in a post -closure safety assess-
ment. This poses a difficult problem for the post-closure safety assessment of radioactive
waste disposal sites, which is commonly based on what may be called the historical as-
sumption-the past is an accurate reflection of the future. However, the geological history
of the earth, over the past few million of years, and the paleontologie record of Homo
sapiens (modem humans), over the ±100 000 years that they exist, indicate that this is not
such a dubious assumption as one may think at first. The real problem of a safety assess-
ment therefore does not seem to be so much the historical assumption, than the approach
used to perform the assessment.
Many interest groups view the post-closure safety assessment of a radioactive waste dis-
posal site as a pure scientific problem that can be solved by the same scientific principles
applicable to a pre-closure assessment. However, as pointed out in Section 1.2, this is not
true. What is needed for a post-closure assessment, is what one can call a compliance ap-
proach rather than the predictive approach of a pre-closure assessment (Kozak, 1997). In
other words, one tries to ensure that the disposal site will satisfy the regulatory criteria at all
times, now and in the future, but not to predict its future. This can often be accomplished
with less effort than that needed by the predictive approach, especially if a pre-closure
assessment of the site has been conducted or is conducted before or during its operational
life.
6.6 FRAMEWORK FOR THE POST-CLOSURE SAFETY ASSESSMENT
6.6.1 Safety Assessment Context
The main purpose of the assessment context is to define the scope and content of the post-
closure safety assessment that is to be performed. Questions that the investigator could ask
during this phase are for example: what are you trying to assess and why are you trying to
assess it? The main effort of the exercise must, however, be to establish a set of higher level
assumptions and constraints that will reflect the regulatory framework, purpose and focus
of the safety assessment, what to include or exclude from the assessment and the justifica-
tion for the choices.
Basic Principles of a Post-closure Safety Assessment 89
As mentioned earlier, the main aim of a post-closure safety assessment is to ensure that
waste management decisions made today do not affect the post-closure safety of the dis-
posal system adversely. Special attention should therefore be paid to waste management
aspects, such as the selection and design of a suitable disposal system, site selection and
characterization, waste acceptance criteria, data collection and remedial actions.
Each regulatory authority will usually impose its own regulatory requirements and princi-
ples on a post-closure safety assessment that must be taken into account in the assessment
context. (See Environmental Agency et al. (1997), for an example.) These may include
aspects such as the principles that should apply to protect the public, regulatory require-
ments (e.g., protection criteria) and more technical requirements such as institutional con-
trol and time frames. In addition, a summarized description of the site, waste and repository
characteristics can be included, because it contains information that may influence the as-
sumptions of the assessment and that should be documented in the assessment context.
Institutional control can be described as the control of a disposal site by an authority or
institution designated under the laws of a country or state (IAEA, 1993a). The control (which
will be an important factor in the design of the disposal facility) may be exercised either
actively (monitoring, surveillance, remedial work) or passively (land use and access con-
trol). Special attention should be paid to the periodes) of institutional control; because it
defines the extent to which humanity can ensure that the waste is safely confined and pre-
vent intrusion into the repository. The institutional control period is normally culture spe-
cific, and may depend on the history of the country. Control periods typically vary from
100 - 300 years (IAEA, 1997a).
The regulatory authority usually sets the time frame that should be used in the assessment.
Since this frame can have a major influence on the assessment results, it may be advanta-
geous to consider other options in the assessment context. One approach would be to extend
the assessment over the full time that the activity of the waste is above the regulatory crite-
ria. Although the activities of all radionuclides decline exponentially with time, the half-
lives of some of them are so long (- 106 years) that the assessment will have to be continued
almost indefinitely, if this criteria is used to establish a time frame. Another approach is to
simply define a cut-off time of say hundreds or thousands of years. The problem with this
approach is that the peak activities at large distances from the near-field will occur much
later than at the repository. Regulators are consequently reluctant to specify a cut-off time.
The most used approach is therefore to perform a post-closure safety assessment until one
can say with reasonable confidence that the peak activities have occurred at all points of the
site's geosphere.
Basic Principles of a Post-closure Safety Assessment 90
The radiological protection criteria-the criteria used to decide whether the proposed dis-
posal system is acceptable or not-is usually set by the regulatory authority. There are
essentially five indicators that can be used for this purpose. These indicators and their asso-
ciated safety criteria are illustrated in Figure 6-3. Which indicator to use depends on the
purpose of the assessment, which also indicates the end entity of the assessment (e.g., a
flux, concentration, and dose). For this reason, they are also known as assessment endpoints
or calculational ene/points.
As pointed out in Section 6.2 a model, more specifically a conceptual model is the only
means with which the behaviour of a disposal system can be studied in some detail over the
time frames of a post-closure safety assessment. Since none of these models is able to
predict the future behaviour of the disposal system with certainty, the uncertainty in the
post-closure safety criteria will increase as one moves from the radioactive waste to its
impact on people's health in Figure 6-3. Despite this uncertainty, the individual dose and
risk indicators remain the most widely used assessment endpoints in post-closure safety
assessments, although the waste activity or concentration is also used sometimes. This is
especially the case when the assessment is aimed at the definition of site-specific waste
acceptance criteria (lAEA, 1999).
The individual dose and risk indicators are measures of the extent that the disposal system
will affect the health of individuals near the facility. To determine these indicators all path-
Indicator Criterion
~ Flux through Engineered Barriers Flux Derived from Engineered Barriers
c ~ _L _
.~ Flux from Geosphere to Biosphere Comparison with Flux of Natural Radionuclides
-
::~> ~==E=n=v=ir=o=nm==en=t=al~~C=o=n=ce=n=tr=a=tio=n===--Comparison with levels of Natural Radioactivity
.S ~~------~-------~ --- -~------- - -
oS
QJ Individual Dose Dose limit and (or) Comparison with Background
cb ~~--------~-----~-----------
Risk Indicator Risl(Constraint
Figure 6-3 The safety indicators used in the post-closure safety assessment of a radio-
active waste disposal site and their associated safety criteria (lA EA, 1997b).
Basic Principles of a Post-closure Safety Assessment 91
ways of radioactive material or radiation from the disposal system to people must be con-
sidered. To determine a suitable criterion is another matter though. The criterion in use
today (a dose limit of 1 mSv·a·l) was established by the ICRP (1990) for real doses to real
people coming from real practices and not for future generations. However, if one accepts
the view that the fourth fundamental principle for the management of radioactive waste
does not discriminate against future generations, a similar criterion can be used here. It is
consequently assumed in post-closure safety assessments that the projected dose to future
members of the public, arising from events likely to occur at the disposal facility, shall not
exceed an appropriate fraction of the present dose limit of 1 mSv a". The appropriate frac-
tion, called the dose constraint, is usually a fraction (0.1 - 0.3) of the dose limit (lAEA,
1999a).
The term risk is usually defined in safety assessments as the product of the probability that an
individual will be exposed to the radiation and the probability that the exposure will give rise
to a detrimental health problem in the individual. The safety criterion commonly used for the
risk indicator states that the predicted radiological risk to individuals, arising from the waste
disposal facility, should not exceed 5.10-5 fatal cancers or other serious generic effects in a
year (IAEA, 1999a).
6.6.2 Description of the Disposal System
The components of the disposal system, the flow of information between them, and the
information that is required to characterize the different components, have already been
discussed in Chapter 4. During the first iteration of a post-closure safety assessment, the
disposal system description is mainly limited to a description of the available information in
the different elements of Figure 4-1. The emphasis will therefore be placed more on the
existing data, rather than the collection of new data. A detailed system characterization is
therefore not required at this stage. However, the emphasis will shift towards a more detailed
site characterization and the collection of new information and data in subsequent iterations.
This may also provide additional information for assumptions and constraints imposed on
the assessment, which may require an update of the assessment context.
As mentioned before, a summarised description of the disposal system can be included
logically in the assessment context, because it contains information that could influence
assumptions to be documented in the assessment context. Together with the assessment
context, the system description will therefore form the basis for the description and justifi-
cations of scenarios in the next step of the safety assessment process, and to a great extent
on the success of the assessment.
Basic Principles of a Post-closure Safety Assessment 92
A considerable effort was often directed towards the collection and interpretation of site
specific data in the past that often led to an over-characterization of the site. Since this is a
time consuming and expensive exercise, it is very important that new data and information
accumulated in subsequent iterations should complement the safety assessment, particu-
larly the regulatory compliance decision. This aspect will be discussed further in Chapter 9.
6.6.3 Description and Justification of the Site's Future Evolution
Radioactive waste has the advantage above chemical and toxic waste that its activity de-
creases with time. However, the long half-lives of many isotopes makes it virtually inde-
structible as far as human life is concerned. The major effort in a post-closure safety assess-
ment is therefore to determine what effect the radioactive waste will have on future genera-
tions and under future conditions, something that is obviously not known.
An approach commonly used to circumvent this problem is to collate and screen all
currently available information on the characteristics of the disposal system (e.g., be-
haviour of people, climate change, geohydrological conditions), and any other natural
or human induced condition. Personal judgement is then used to develop a number of
. scientifically sound descriptions, commonly referred to as scenarios, of the future con-
ditions at the site.
According to the Oxford English Dictionary the word scenario (like model) has so many
meanings and is so overused that it is not always possible to determine its precise meaning.
However, in this thesis the word will be used exclusively in the following sense.
A scenario is one member in a set of hypothetical features, processes and events devised
to rationalise the future behaviour of a repository system in a safety assessment.
This definition is very similar to the one given by Chapman et al. (1995), based on their
experience in the SKI SITE-94 study, in which special attention was given to the development
of scenarios through formal techniques. Other definitions for the word can be found in
lAEA (1999b).
The scenario approach has the advantage that it allows for a mixture of quantitative analysis
and qualitative judgements of the future conditions. In principle, a number of scenarios can
be generated. In practice, however, a base case or normal scenario is selected, followed by
a selection of alternative and disruptive scenarios that are most likely to have the greatest
impact on system performance. Care should, nevertheless, be taken to ensure that the se-
lected scenarios provide an appropriately comprehensive picture of the system, its possible
Basic Principles of a Post-closure Safety Assessment 93
evolutionary pathways, critical events and system robustness, based on the assessment con-
text and system description. The transparent generation of the scenarios and their associ-
ated justification methodology can be very important in building public confidence for the
post-closure safety assessment. The scenarios and the methods used to generate them are
also often the main focus points of independent reviewers of the safety assessment (!AEA,
1997a). More details of the procedures that can be used to develop and justify scenarios can
be found in Chapter 8.
6.6.4 Formulation and Implementation of Models
Once the scenarios have been developed, their consequences must be analysed in terms of
the assessment context. Although a qualitative analysis may be sometimes sufficient, there
are situations where a quantitative analysis will be necessary. The most direct approach to
achieve this is to develop a conceptual model for each of the scenarios, as discussed in
Section 6.2. The mathematical models, on which the conceptual models are based, can be
derived from either sound physical principles or empirical observations, depending upon
the level of understanding of the site and the available information
It will seldom be necessary to develop new mathematical models for the scenarios encoun-
tered in safety assessment analyses. These scenarios are usually closely related to subjects
such as: groundwater flow, heat flow and air dispersion, for which there exist a large number
of mathematical models. Only if interactions are identified that are not included in the es-
tablished models, will it be necessary to develop new mathematical models. The only situ-
ation where this may be necessary is when observations indicate that unknown or less un-
derstood interactions influence a decision adversely.
As explained in Section 6.2, a conceptual model can be expressed either in the form of an
analytical model or a numerical model. Numerical models and in some cases analytical
models are today often implemented in commercial or private computer packages which
are widely available. Unfortunately, many of these packages do not describe the underlying
mathematical model in detail. The possibility therefore exists that a safety analyst may use
a computer package that does not satisfy his or her conceptual model for a scenario. It is
therefore very important that the safety analyst should ensure that the mathematical model
used in the chosen computer package is consistent with the conceptual model of the sce-
nario that has to be investigated. One must also keep in mind that in some cases it will not
be desirable, nor feasible, to model a scenario with a single computer package. More details
of the procedures that can be followed to develop conceptual models for each scenario can
be found in Chapter 8.
Basic Principles of a Post-closure Safety Assessment 94
The level of detail needed in a conceptual model and its associated mathematical model will
depend, not only the assessment context, but also the stage of iteration of the safety assess-
ment. It may therefore be possible to begin an assessment with simple analytical models
and then proceed to more complex numerical models as the assessment proceeds.
An important question to ask in choosing a conceptual model for the assessment of a radio-
active disposal system is, how will the uncertainties be handled? For example, will a deter-
ministic model (perhaps adjusted for statistically distributed constitutive parameters and
boundary conditions) be sufficient, or does one need a full probabilistic model? This will
have a significant implication on the way the rest of the assessment will be done and should
therefore be defined in the assessment context. Although some regulatory authorities may
provide guidance in this connection, it may not be a bad idea to use both types of models
during the first one or two iterations of the investigation. Attention should also be given to
the type and quantity of data needed for the model. The data collected during the initial
characterization of the disposal system, and generic data may be sufficient for the first few
iterations steps, but latter iterations may require more or other types of site-specific data. It
therefore makes sense to make sure that the computer package(s) used in the modelling can
accommodate these additional data.
6.6.5 Consequence Analysis
Once the necessary models are formulated, the major purpose of the modelling exercise is
to determine what consequence will a particular scenario have on the future development of
the disposal site-hence the name consequence analysis. For this purpose, the definition in
Section 6.3 is very important. This emphasizes that a safety assessment analysis is a pro-
spective evaluation. Not with the intent to predict its actual behaviour in the future, but
rather to understand the behaviour of the system better and to reflect the importance of
specific components with respect to the compliance criteria.
It is important to ensure that the consequence analysis results are consistent with the assessment
context, for example, that the correct endpoints are calculated over the time frame of concern.
Depending on the assumptions made in the assessment context and the models developed during
the previous stage, the analysis will be done in a deterministic or probabilistic mode or both.
Incorporated into the consequence analysis, is uncertainty and sensitivity analysis. Uncer-
tainty analysis can be defined as an estimation of uncertainties and error bounds of the
quantities involved in and the results from the solution of a problem (IAEA, 1993a). This
requires the application of statistical techniques and definition of the input data information
Basic Principles of a Post-closure Safety Assessment 95
in probabilistic form. Uncertainty analysis will be discussed in more detail in Chapter 8.
Sensitivity analysis is a quantitative examination of how the behaviour of the system varies
with changes in the values of the initial conditions, boundary conditions, parameters built
into the code and parameters that must be provided as input data. Attention should in addi-
tion also be paid to how sensitive the results are to the general assumptions on the system
behaviour and whether all the considered processes needs to be included in the model de-
velopment or not (IAEA, 1993b). Two common approaches often used in sensitivity analy-
sis (IAEA, 1993a) are:
(a) Parameter variation, in which the variation of the results is investigated for changes
in one or more input parameter values within a reasonable range around a selected
reference or mean value, and
(b) Perturbation analysis, in which the variations of the results with respect to changes
in all the input parameter values are obtained by applying differential or integral
analysis.
It is clear from Figure 4-1 that a consequence analysis will be required for one or more of
the near-field (engineered barriers), the geosphere (natural barriers) and the biosphere. Which
component(s) to include in the analysis will dependent very much on the type of assess-
ment, that is whether it is a full safety assessment, or just to investigate the performance of
one component. Consequence analysis forms therefore an integral part of any post-closure
safety assessment, as shown by the more detailed discussion in Chapter 7.
6.6.6 Interpretation of Results
The first opportunity a safety analyst will have to interpret the results is after the modelling
of at least one of the different scenarios has been completed. However, a meaningful inter-
pretation of the assessment in terms of the assessment context can only be done after a first
iteration of all scenarios. At this stage he or she will have quantitative data available to
compare with the regulatory and other criteria, outlined in the assessment context, to assess
reasonable assurance of compliance.
The interpretation of data is a multi-faceted process in that several varied and sometimes
competing factors must be brought together and reconciled to reach a decision as to whether
the system and post-closure safety assessment are adequate. A major activity will therefore
be to screen and condition data from the modelling results (even eliminate unnecessary
data) before comparing it with the relevant criteria. Once this is done, the analyst will be
Basic Principles of a Post-closure Safety Assessment 96
able to make a decision.
Since the post-closure safety assessment is an iterative process, the process will usually
follow the branch to the second decision process in the flow chart of Figure 6-1, if compli-
ance cannot be demonstrated after the first iteration step. In this case, a second iteration of
the safety assessment is appropriate, which-according to Figure 6-1-requires an assess-
ment for further information needs. The information that is required will be determined by
the reason for the inadequacy, although it can. be expected that it will be those input param-
eters and assumptions that have the greatest impact on system performance. Once the infor-
mation needs have been identified, the next step is to assess the worth of the data associated
with collection of data to resolve the input parameter and model assumption inadequacy.
The assessment of data worth will again be discussed in Chapter 9. If subsequent iterations
show that it will be inefficient to further try to improve system components and the system
still does not comply with the regulatory criteria, the system must be rejected.
6.6.7 Confidence Building
As mentioned in Section 4.3.2, it will most likely be impossible to develop a disposal a site
for the disposal of radioactive waste without the confidence, co-operation and participation
of the local population and other interested parties. Scientists and engineers working on the
technical aspects of radioactive waste disposal have developed an international consensus
that the waste can be permanently managed in a manner that protects the environment and
public health. However, this view is not necessarily shared by the public (IAEA, 1992). A
post-closure safety assessment must therefore be conducted in such a way that this confi-
dence building is always enhanced. This may be achieved by open discussions with all
interested parties (stakeholders) in the early stages of the assessment, supplemented later
with discussions of the site-specific results as the safety assessment proceeds. Apart from
the general public, other stakeholders include national and provincial governments, regula-
tory agencies, local communities, the media, environmental groups, and employees in the
nuclear industry and opinion leaders.
Itmay also be helpful to demonstrate to the interested parties that the assessment is based
on sound scientific and engineering principles. For this purpose, efforts are directed to-
wards three areas, namely model verification and validation (natural analogues), quality
assurance, and critical peel' review.
Verification is the process of showing that a mathematical model, or the corresponding
computer code, behaves as intended, i.e. that it is a proper mathematical representation of
----------------------------- "_
Basic Principles of a Post-closure Safety Assessment 97
the phenomena it represents and that the equations are correctly encoded and solved. Vali-
dation, on the other hand, is a process carried out by comparison of model predictions with
field observations and experimental measurements (IAEA, 1993a). A model is considered
validated when it is able to simulate a sufficient number of observations on the phenomenon
for which it was developed. Natural analogues (Apted and Engel, 1991) may be especially
useful in this regard. These are features similar to that investigated in the safety assessment,
but occur naturally in nature for very long times. The confidence in a site assessment model
will certainly increase considerably if it is able to simulate its natural analogue counterpart
accurately (!AEA, 1993a). Most of the known natural analogues are associated with physi-
cal and chemical processes in the geosphere, such as the migration of radioactive materials
and other deposits and matrix diffusion. However, there are others relating to the near-field
and biosphere. These include, for example, corrosion and penetration of waste containers,
waste form degradation, stability of backfill materials, elemental solubility, sorption proc-
esses, microbial activity and speciation (IAEA, 1992).
Quality assurance comprises all those planned and systematic actions necessary to provide
adequate confidence that a structure, system, process or component will perform satisfac-
tory (!AEA, 1993a). Quality assurance is normally applied to repository design and site
characterization. However, the need to generate confidence in safety assessments dictates
that appropriate quality assurance methods is also applied to aspects such as data collection,
model development, computer code development and integrated assessments. These meas-
ures will contribute significantly to reducing uncertainty and increasing confidence in safety
assessment results (!AEA, 1993b). The treatment of uncertainties in safety assessment will
be discussed in detail in Chapter 8.
Critical peer review forms an integral part of the scientific process to gain acceptance of a
new concept. This allows a scientist to publish his results in the open literature where it can
be reviewed and criticized by other experts in the field. It is therefore logical to assume that
a safety assessment can also benefit from such a critical peer review. Such a peer review
should preferably be done by a group of people who are experts in the field, but not associ-
ated with the assessment to be reviewed. The major task of such a group will be to review
the assessment critically and make suitable recommendations (IAEA, 1993b).
CHAPTER7
PROSPECTIVE EVALUATION OF A DISPOSAL SYSTEM
7.1 INTRODUCTION
The assessment of a radioactive waste disposal site was historically often seen as an attempt
to predict the behaviour of the site far into the future, using deterministic or predictive
models. The models, which are discussed in Section 6.2, are based on the physical princi-
ples that underlie a specific phenomenon and should therefore be able to predict the future
behaviour of the site. The models, unfortunately, frequently require information on param-
eters whose behaviour cannot be determined with certainty far into the future. It is therefore
impossible to apply the models in the historical sense of a safety assessment. However, as
pointed out in Section 6.3, this is neither necessary nor desirable. What one really wants to
do in a safety assessment is to be able to give a reasonable assurance that the site will
comply with the regulatory criteria for safety. This led to the introduction of the term com-
pliance calculations (Kozak, 1997).
Compliance calculations in a safety assessment essentially involve the prospective evalua-
tion of radionuclide transport through the near-field, geosphere and biosphere, where peo-
ple will be exposed to the detrimental effects of the radiation. As shown by the discussion of
the near-field and geosphere in Sections 7.2 and 7.3 respectively, this can often be achieved
by rising predictive models, but applied in a heuristic and not predictive sense.
Heuristic predictive models can also be used to evaluate physical structures in the bio-
sphere, such as wind erosion of land-fills, but not the biodynamical components of the
biosphere (food chains, for example). These components are best evaluated with the
phenomenological models that are described in Section 7.4.
7.2 THE NEAR-FIELD
7.2.1 General
The definition of the near-field in Section 4.3.4 indicates that it is the primary source of
radiation to the environment at a waste disposal site, and at the same time the first defence
against the leakage of radioactive nuclei from the site. It must therefore be considered as the
-98-
Prospective Evaluation of a Disposal System 99
most important component in the safety assessment of a radioactive waste disposal system.
The evaluation of the near-field in a safety assessment context involves the characterization
of the groundwater and vapour motion, especially if the repository is situated in the unsatu-
rated zone, and the associated transport of radionuclides. The mass flux of radionuclides
across the boundary of the near-field is thus the primary source of radiation at a waste
disposal site.
Radionuclides can escape from the near-field in either one of two phases-a gas phase and
a solid phase. The gas phase will ultimately escape to the atmosphere where its transport
can be studied with an atmospheric transport model (Ellis, 1999). This discussion, however,
will concentrate exclusively on the groundwater pathway in the geosphere, although it is
recognized that the atmospheric pathway may be important, especially for near-surface
disposal systems.
The near-field characteristics of a near-surface disposal system will differ considerably from
that of a geological disposal system, because of differences in inventory, engineered barri-
ers and design of the repository. The two systems will therefore be discussed separately.
7.2.2 Near-Surface Disposal Systems
The rate at which radionuclides will leak from the near-field will be determined essentially
by the degradation of the containers and engineered barriers, and the subsequent distribu-
tion of the nuclides in the near-field. Although there are other factors that may cause the
containers and engineered barriers to fail, the metal and concrete used in the conditioning of
the waste and engineered barriers are especially susceptible to infiltrating water. Some of
the processes responsible for the degradation of these structures by infiltrating water are
listed in Table 7-1. Phillip and Clifton (1989) also provide a good overview of the proc-
esses, including estimates of the service life of different concrete structures.
The strength of near-field source, caused by the infiltration of water, will be determined by
a number of factors, of which the following four are the most important (Sullivan 1997):
(a) the inventory bywaste form and container,
(b) container performance,
(c) waste form performance, and
(d) transport of the contaminants through the disposal facility.
Prospective Evaluation of a Disposal System 100
Table 7-1 Degradation processes in engineered barriers associated with the infiltration
of water, taken from Andrade (1997).
Process Mechanism Agent
Biological Expansion, leaching Bacteria
Electrochemical Corrosion Electrochemical potentiaIs
Mechanical Cracking Shrinkage, deformations and loading
Physical Expansion, erosion Freezing and thawing, salt crystallization,
delamination
Physico-chemical Leaching, hydrolysis, Acids, sea water, ammonium salts, sulphates, alkali-
expansion aggregate reactions
The inventory, taken as the quantity of radioactive material in the waste, is obviously the
source of all the radiation that will be released in the geosphere. There are essentially two
approaches that can be used to determine this quantity. The first is on a container or package
basis and the second as a site average. The site-average approach is considerably simpler
than the per package approach and is the one commonly used in site assessments.
Although the inventory is the major source of the radionuclides, the source strength will be
determined by the performance of the container and waste form, that is the abilities of the
container and waste form to limit the migration of the nuclides from the source. An intact
container will usually be able to shield the waste completely from the infiltrating water,
with the result that one has only to take degraded containers into account when estimating
the source term. This can be a very time consuming and expensive exercise, judging from
the time-dependent nature of the processes responsible for the degradation of containers
and engineered barriers in Table 7-1. Such a computation is, however, rarely necessary or
desirable in safety assessments, unless a more simplified approach has a negative impact on
the regulatory decision.
The ability of the waste form to retain radionuclides will be determined mainly by the types
of wastes in the repository and the distribution of the radioactive nuclides in the waste form.
Based on the knowledge of the inventory, the analyst should attempt to define the leaching
mechanisms appropriate for each waste type. Sullivan (1997) defined four models for the
release of radionuclides from the waste form: rinse release with partitioning, diffusion re-
lease, uniform degradation release and solubility-limited release.
(a) Rinse Release with Partitioning - The radionuclides are assumed to be situated on
the surface of the waste form and will be released as soon as the waste form is
infiltrated by water, subject to equilibrium partitioning and solubility limits. Rinse
release without partitioning therefore means that the entire inventory will be re-
Prospective Evaluation of a Disposal System 101
leased instantaneously, once a container fails.
(b) Diffusion Release - The waste is assumed to be uniformly distributed throughout
the waste form and can only escape through molecular diffusion.
(c) Uniform Degradation Release - The waste form releases the radioactivity at a
uniform rate. Uniform degradation release is therefore similar to rinse release with-
out partitioning, except that the radionuclides are released at a uniform rate and not
instantaneously. An example of a situation where this type of release can occur, is
when the container has corroded completely and the waste form itself decays.
(d) Solubility-limited Release - Radionuclides are instantaneously released to solu-
tion until the solubility limit is reached. Further releases are controlled by the mi-
gration of radionuclides away from the waste form.
Radionuclides are mainly transported from the waste in near-surface facilities by the mo-
tion of ground water through adveetion and dispersion. However, there are also other mecha-
nisms, such as diffusion, solubility, sorption and radioactive production and decay that will
contribute to the transport. The contributions of the different mechanisms, however, may
vary with sité and waste type. For example, diffusion will be the dominant mechanism in a
site with intact containers.
It will clearly not be easy to establish a suitable conceptual model for the transport of
radionuclides within and out of the near-field, unless there is sufficient experimental data,
although the situation may vary with site and waste form. Solid concrete, for example, is a
porous material. Its moisture retention and transport properties can therefore be character-
ized with parameters similar to that used in groundwater flow models (Shuman and Rogers,
1992). However, such a model may have to be replaced by a fracture flow model, once the
concrete begins to crack (Walton et al., 1990; Walton and Seitz, 1991). The transport of
radioactive nuclides in the near-field is consequently often modelled with less general, but
more conservative models in safety assessments of near-surface disposal facilities. The time-
delay model, which assumes that all the nuclides, will be released simultaneously but only
after a certain time, is an example of such a model.
7.2.3 Geological Disposal Systems
As mentioned in Section 3.2.3, geological disposal systems are mainly aimed at the dis-
posal of high-level radioactive wastes in deep geological formations, although there is now
a tendency to dispose low- and intermediate-level waste in similar structures. The exact
construction of the repositories used in these systems will therefore differ considerably,
Prospective Evaluation of a Disposal System 102
depending on the design criteria, and whether the repository is situated near, or far below
the surface. A schematic illustration of such a facility is shown in Figure 7-1.
Geological disposal systems are usually constructed with multiple engineered barriers, as
shown in Figure 7-1. Since the main aim with the barriers is to prevent the release of
radionuclides as far as possible, the number included may vary from one to another.
A geological repository will be drained and ventilated during its operational life. It is there-
fore unlikely that radionuclides will be transported from it during this time. However, the
conditions in the near-field will change gradually with time once the repository is closed.
The ensuing 'normal conditions', can be divided in a number of distinct stages (Chapman
and McKinley, 1987).
(a) Water will first infiltrate into the zones of the host rock, drained during the opera-
tional phase, and then the backfill material, causing swelling and plastic flows in
some materials containing swelling clays, thereby sealing constructional gaps.
(b) Radiogenic heat, generated especially by high-level waste, will cause a thermal gra-
dient to develop within a few years after the closure of the facility that may last for
Natural Soil Cover WaterEngineered Cap
1------····.· _.
(Low-level Waste Vault)
I Thermal gradient changes I Radiolysis
2 Heat generation and transport r--:T=h:-"e-r-m-a-7-"-"-1-"I./V'v----,-:-:-::-:-:=---::---;--, 2 Changes in material
3 Mineralogical phase change Field -«--- r---'" characteristics
4 Viscosity changes 3 Decay gas generation
5 Fluid migration 4 Mineralogical changes
ister
Wall or
Disposal Room (including backfilling)
I Leaching
I Stress building
2 Compression Thermome- Chemical 2 Solubility
3 Corrosion
3 Spalling and displacement chanical Field Field
L.....:.:c.=:.:..:..::c..:c..;..:;.;..:;----"v'\iAr----....:....:..:;.;..:;----' 4 Gas generation
4 Microfracturing
Hydrological 5 Geochemical changes
Field
Natural Rock I Hydrothermal degradation of rock-glass system2 Increased permeability
3 Trans of radionuclides
Figure 7-1 Schematic partial cross-section through the wall of a chamber in a hybrid
geological disposal system, with the different processes and components that
could influence the performance of the system's near-field. [Redrawn from
JAEA (1992).]
Prospective Evaluation of a Disposal System 103
thousands of years. This will heat the engineered barriers and the surrounding host
rocks, thereby disrupting the normal groundwater flow pattern in the geosphere of
the facility.
(c) As groundwater infiltrates the backfill, its chemistry tends to change due to a complex
series of reactions, such as ion-exchange, dissolution and precipitation of particular min-
erals. The degree of change will depend on the type of material used as backfill.
(d) The infiltrating water will ultimately reach the canister and begin to corrode it. The
low seepage velocity of the infiltrating water will be not only limit the corrosion of
the containers, but also restrict the mass transport at this stage to diffusion. The
containers, nevertheless, will fail sometime, either because of corrosion or mechanical
forces. However, this does not mean that radionuclides will be released immediately
to the water, since the remnant canister material and corrosion products may still
retard the release of the radionuclides.
(e) The waste form begins to disintegrate when it is exposed to the water, release the
radionuclides, either in solution, as particulates or colloids, and begin to migrate
through the near-field. The rate at which the radionuclides are released into solution
is determined by the solubility of the radionuclides and the rate at which the matrix
disintegrates. The chemical environment in the contact zone is very complex and
difficult to evaluate.
(f) Once released from the near-field, the rate of migration of radionuclides though the
geosphere is determined by the geohydrological conditions along the flow path and
the extent of retardation processes.
7.3 EVALUATIONOFTHE GEOSPHERE
7.3.1 General
The radionuclides released by the near-field to the geosphere pose no risk to humans until
they are released in the biosphere. Special attention should therefore be given to the evalu-
ation of the geosphere as a natural barrier for the transport of radionuclides from the near-
field to the biosphere. The approach commonly used for this purpose is to use the informa-
tion gathered during the site's characterization and the mathematical models for groundwater
flow and mass transport in Section 6.2, to develop a suitable conceptual model for the
geosphere of the site. Of particular importance in this regard are the physical principles
embodied in the mathematical models, that are discussed below.
Although there already exists a large volume of literature on the subject, the time scales
Prospective Evaluation of a Disposal System 104
involved and the amount of data available often introduce considerable uncertainties into
the approach. Several countries have consequently developed underground research labo-
ratories to obtain more site-specific data and improve the various mathematical and concep-
tual models needed for the approach (Kickmaier and McKinley, 1996). Conservative as-
sumptions and analytical models are also sometimes used to compensate for the uncertain-
ties, especially in preliminary safety assessments. However, site-specific data and models
will be needed to license a particular disposal system (IAEA, 1993b).
7.3.2 Adveetion
Radionuclides in the geosphere are predominantly transported with the water present in the
pores or fractures of the geosphere matrix-a process known as advection. The advective
transport of the nuclides, represented by the term q. Vc, in Equation (6.4), is mainly control-
led by the concentration gradient of the dissolved radionuclides and the Darcy velocity of
the water. Since the Darcy velocity differs considerably from an unsaturated to a saturated
medium, the degree to which the geosphere matrix is saturated with water can play a signifi-
cant role in the transport of the radionuclides by the water.
The flow of groundwater in consolidated rocks is often completely controlled by fractures
(Botha et al., 1998). One therefore has to replace the simple porous flow models of Section 4.4
with more complex mathematical models that takes the aperture, direction and frequency of the
fractures into account, when studying these formations. It is consequently not an easy task to
predict travel times, flow rates and contaminant discharge points in these formations.
7.3.3 Mechanical Dispersion
The term mechanical dispersion is a collective term used to describe microscopic processes
that cause dissolved solids to mix and spread in groundwater. It is generally assumed that
the phenomenon is caused by interactions between the dissolved solids, pore structure and
the microscopic velocity of water in the pores (Freeze and Cherry, 1979), although very
little is known of the exact mechanisms responsible for the phenomenon.
The main effect of mechanical dispersion in ground water is to spread the contaminants over
a larger volume than would have been the case if dispersion were absent, and in the form of
a plume. This tends to reduce the peak concentrations of contaminants at the front of the
plume, but at the same time increase the concentrations on its sides.
There do not exist methods with which the generalized microscopic Darcy velocity can be
Prospective Evaluation of a Disposal System 105
measured routinely (Botha, 1996). Mechanical dispersion is therefore usually expressed in
terms of the macroscopic quantities, such as the seepage or Darcy velocity of the fluid,
through the second rank tensor, Oh' in Equation (6.6). This representation has an important
consequence for groundwater pollution studies, as it implies that the main axis of a con-
tamination plume may not coincide with the direction of the macroscopic Darcy or seepage
velocities. The behaviour, which is particularly noticeable in heterogeneous aquifers, unfor-
tunately, is often disregarded in studies of groundwater pollution.
7.3.4 Molecular Diffusion
It is well-known that a dissolved solid will tend to spread evenly though the fluid, even if the
fluid is at rest. The random thermal motion of molecules in the solution causes this mecha-
nism, known as molecular diffusion, and represented by the molecular diffusion tensor in
Equation (6.5). This mechanism is therefore always present in all fluids and solids, except
at the zero-point temperature (0 K).
The major effect of molecular diffusion is to transport solids from areas with high concen-
trations of a substance to areas with low concentrations of the same substance. Molecular
diffusion depends only on the diffusing substance, the medium in which it is embedded and
its concentration gradient within that medium. It can therefore occur even in stationary
ground water or between groundwater and the surrounding rock matrix.
The first impression of molecular diffusion is that it will retard the motion of contaminants
through the subsurface of the earth. Because of this and the fact that the molecular diffusion
coefficient for ordinary substances in water is not very large, molecular diffusion is often
neglected in groundwater pollution studies (Botha, 1996). However, such an approach can
have dire consequences, not only for radioactive waste disposal systems, but also for all
waste disposal systems, for a number of reasons. The first is that molecular diffusion is not
restricted to fluids alone, but can occur in any form of matter. This means that no engineered
barrier, even those that may last for thousands or millions of years, will ever be able to
prevent radionuclides to escape from the near-field. A second reason is that molecular diffu-
sion will tend to broaden the adveetion plume further, even contaminate adjacent aquifers
separated by an impermeable flow boundary from the contaminated one.
To illustrate the latter behaviour, consider the two-layered heterogeneous aquifer system in
Figure 7-2. The higher hydraulic conductivity of Layer 2will obviously cause this layer to
form a preferential flow path in the aquifer, along which adveetion of pollutants will domi-
nate. However, this will create a concentration gradient between the contaminants in Layer 1
Prospective Evaluation of a Disposal System 106
Molecular Diffusion Layer I
/
Advcelion
Molecular Diffusion Layer I
Figure 7-2 Schematic illustration of adveetion and molecular diffusion in a two-layered
heterogeneous aquifer, after Botha (1996).
and those in Layer 2. This gradient will not only cause the pollutant to diffuse into Layer I,
but also decrease the concentration of the pollutant in Layer 2, thereby retarding the pollut-
ant's motion and at the same time smooth out any variations in the concentrations of the two
layers. Studies of contaminant transport, over the last couple of decades, have shown that
molecular diffusion can often dominate adveetion in some aquifers, because of this behaviour.
This is especially the case in heterogeneous and fractured rock aquifers (Sudicky, 1983).
7.3.5 Retardation Processes
Molecular diffusion is not the only mechanism able to retard the transport of a dissolved
solid in a porous or fractured medium. Another very important example is sorption. Al-
though sometimes regarded as a single phenomenon, there are quite a number of different
mechanisms that can contribute to sorption. Some of the most important of these are:
(a) absorption of the dissolved molecules by the rock matrix,
(b) precipitation of the dissolved molecules caused by changes in the pH-value of the
groundwater,
(c) ion exchange and the isomorphic substitution of ions between the solute and the
rock matrix,
Although it is possible to account for all these mechanisms in the conceptual model of a
waste disposal site's geosphere, the existing geochemical databases are often not sufficient
for this purpose. Moreover, there are not many computer resources in the world that can be
used to evaluate such a conceptual model. Almost all safety assessments of radioactive
waste disposal systems are consequently limited to conceptual models in which the sorp-
tion mechanisms are presented by a single parameter, K", known as the distribution coeffi-
cient or simply K,(va/ue, defined by the equation
Prospective Evaluation of a Disposal System 107
s(x,t) = Kdc(x,t)
(also known as the Freundlich isotherm) where sex, t) is the mass fraction of the solute that
partakes in sorption, and c(x, t) the volumetric concentration of the solute. Substitution of
this value for sex, t) into Equation (6.4) yields :
where
n, =()+ PbKd
is known as the retardation coefficient.
The distribution coefficient represents a very efficient method with which to study sorption
in the framework of a safety assessment, despite its simplistic nature. It is very important
though that the Kd -values used, should be representative of the site. Too high Kd-values will
tend to concentrate the radionuclides in the vicinity of the near-field, while too low Kd -
values will tend to rinse the near-field of radionuc1ides, with adverse effects for other com-
ponents of the assessment. Too high Kd -values can, for example, lead to an overestimate of
the dose humans will receive on intrusion of the site, while members of a radioactive chain
will not have sufficient time to produce daughter elements in the second case. The latter
situation should especially be avoided in the case of elements, such as mU, whose daugh-
ters (226Ra,222Rnand 21OPb)are often more active than the parent (Sullivan and Kozak,
1996).
7.3.6 Radioactive Decay
Radioactive decay itself will reduce the activities of the radionuclides in the geosphere with
time. Unfortunately, the half-lives of some of the radionuclides in radioactive wastes are too
long for their activities to be reduced noticeably during the time they are transported through
the geosphere. One cannot therefore rely on radioactive decay alone to reduce the activities
of all radioactive waste to acceptable limits in the far future.
A second difficulty with radionuclides is that the daughters of elements in decay chains are
completely different elements, in general, with completely different physical and chemical
properties, and therefore transport properties. Although possible, it is not always easy to
provide for these daughters in the conceptual models used in the safety assessments of
radioactive waste disposal systems. One approach to avoid this is to assume that the parents
and daughters are always in radioactive equilibrium. Since the rate at which a parent pro-
duces a daughter is known, one only needs to compute the activity of the parent and used
Prospective Evaluation of a Disposal System 108
this to compute the activities of the daughters. This approximation yields excellent results
when the half-lives of all daughters are considerably less than that of the original parent,
and the transport rate is slow enough to ensure equilibrium.
Another approach that is also used is to assume that the parent and daughters are trans-
ported in the same manner and at the same rate. One then only need to compute the activity
(c.) of the parent in the geosphere and use the rate at which the i-th daughter is produced by
J
the j-th parent and the generalized Bateman equations (Skrable et al., 1974)
to compute the activities of the daughters (c) as a function of time.
7.4 EVALUATIONOF THE BIOSPHERE
The transport of radionuclides through the biosphere are better understood than through the
natural and engineered barriers (IAEA, 1993b). The major uncertainties occur in the evolu-
tion of the biosphere in the future and the changing patterns of human behaviour.
It follows from Figure 7-1 that the results from the evaluation of the geosphere and the
characteristics of human behaviour form the basis for the evaluation of the biosphere. Two
kinds of models are conventionally used to evaluate the behaviour of the biosphere-deter-
ministic and phenomenological models.
Deterministic models are conventionally used to describe mechanistic phenomena, such as
wind erosion of land-fill and soil surfaces, while the phenomenological models are used for
the biological pathways, such as the food chain. The phenomenological models used in
radioactive safety assessments are usually based on transfer coefficients. Although these
models are not able to explain a phenomenon, as is the case with deterministic models, they
are simple to use and based on phenomenon specific data (Torres and Simon, 1997).
The main purpose with models of the biosphere is to estimate the radiation doses that peo-
ple will receive when exposed to the environment. The potential pathways that may cause
such an exposure are usually divided into two groups: internal and external. The internal
pathways consist of the ingestion of contaminated food; the drinking of contaminated water
and the inhalation of contaminated air. The external pathways, which are more diffused
include exposure to contaminated ground, houses built with contaminated materials and
swimming or bathing in contaminated water. These pathways are usually divided into 'com-
Prospective Evaluation of a Disposal System 109
Table 7-2 Parameters and equations often used to compute the radiation dose received
by humans through pathways in the biosphere of a radioactive waste dis-
posal site.
General parameters
DFing = Dose conversion factor for ingestion Sv Bq-I
Cs = Concentration in soil Bq kg-I
Cw = Concentration in water Bq L-I
CMMxxx = Annual individual consumption L k_g_
Drinking water dose Dw
Dw = DFing . Cw . CMMwater Sv a:'
Dose from milk consumption Dmi
Dmi = DFing . CMMmilk . DFmilk . A Sva-I
DFmilk = Transfer factor for milk d L-I
A = CMCgrass . Cs + CMCwater . Cw -
CMCgrass = Daily grass consumption by cows Kgd-I
CRgrass = Concentration ratio from soil to grass -
CMCwater = Daily water consumption by cows L d-I
Dose from meat consumption Dme
Dme = DFing . CMMmeat . DFmeat . A Sva-I
DFmeat = Transfer factor for meat dkg-I
A = As under 'dose from milk consumption' -
Dose from consumption of leafy vegetables Dbl
CRIeaf = Concentration factor from soil to leafy vegetables -
Dbl = DFing . CMMleaf . CRleaf . Cs Sva-I
Dose from cereal consumption Dget
Dget = DFing . CMMcereal . CRcereal· Cs Sva-I
CRcereal = Concentration factor from soil to cereal -
Dose from consumption of root vegetables Dwu
Dwu = DFing . CMMroot . CRroot . Cs Sva-I
CRroot = Concentration factor from soil to root vegetables -
Dose from fruit consumption Dft
Dft = DFing . CMMfruit . CRfruit . Cs Sva-I
CRfruit = Concentration factor from soil to fruit -
Dose from poultry consumption Dpt
Dpt = DFing . CMMpoultry . DFpoultry . B Sva-I
DFpoultry = Transfer factor for poultry d kg:'
B = CMHcereal· Ckcereal- Cs + CMHwater· Cw -
CRcereal = See 'dose from cereal' -
CMHcereal = Daily cereal consumption by hens Kgd-I
CMHwater = Daily water consumption by hens Ld-I
Dose from egg consumption Dei
Dei = DFing . CMMegg . DFegg . B Sva-I
DFegg = Transfer factor for eggs dkg-I
B = See 'dose from poultry consumption' dkK'
Dose from fish consumption Dfi
Dfi = DFing . CMMfish . CRfish . Cw Sv a-'
CRfish = Concentration factor from water to fish -
Prospective Evaluation of a Disposal System 110
partments' in which the radionuclides are distributed instantaneously and homogeneously
in biospheric models. The models also assume that the rate at which the radionuclides are
transported from one compartment to another is proportional to their concentration in that
compartment. The transport of the radionuclides between the various compartments (e.g.
water to grass, grass to cow, cow to milk or beef, milk or beef to man) is then modelled with
two simple parameters the transfer coefficient and concentration ratios (CR). The dose re-
ceived by a person is finally computed from dose conversion factors, as illustrated in Table
7-2. These factors are based on the experience with radioactive materials over the last cen-
tury and make provision for variations of dose per unit intake, the age of the individual and
the chemical properties of the radionuclide (IAEA, 1995b).
CHAPTERS
UNCERTAINTIES IN SAFETY ASSESSMENTS
8.1 INTRODUCTION
The discussion of the principles and the evaluation of safety assessments in Chapters 6 and
7 may create the impression that this is a 'squishy' problem. That is, the analysis appears to
be rigorous on the surface, but are underlain by large and irreducible uncertainties (Kozak,
1994). Safety assessments must extrapolate over large spans of time and space, into condi-
tions that cannot be observed empirically. Moreover, the currently prevailing conditions at a
site are often so complex that it confounds all attempts topredict the actual behaviour of the
system accurately. However, as mentioned in Section 1.2, the main aim of a safety assess-
ment is to determine the extent to which a waste disposal site will comply with the regula-
tory constraints and not to predict its actual future behaviour. The uncertainties inherent in
quantitative post-closure safety assessments thus mean that results are to be considered as
representative indicators of safety, rather than definitive predictions of the future radiologi-
cal impact (Watts et al., 1999). A safety assessor is therefore not so much concerned with
numerical uncertainties in the final results of the different models used in the assessment, as
') is customarily the case when using models to describe physical phenomena. What is impor-
tant here is whether changes in the assumptions, models and parameters used in the assess-
ment can affect the decision on the regulatory compliance of the disposal system adversely
(Kozak, 1997). There is consequently a vast difference between a numerical uncertainty
analysis as practised in many branches of science and engineering, and uncertainty analysis
in the assessment of radioactive waste disposal sites. The Scientific Committee 87-3 of the
NRCP therefore introduced the term Importance Analysis in an attempt to clarify the differ-
ence between the two approaches, and to stress the fact the processes associated with a
safety assessment are themselves uncertain. However, this does not change the fact that
uncertainties do exist in safety assessments, and that they must be addressed, before anyone
will accept an assessment with confidence. In fact, the treatment of uncertainty is central to
the establishment of the post-closure safety case for a radioactive waste disposal system in
the United Kingdom regulatory guidelines (Environment Agency et al., 1997).
The uncertainties associated with safety assessments can be conveniently divided into three
categories:
(a) uncertainties related to the unknown future state of the disposal system,
-111-
Uncertainties in Safety Assessments 112
(b) data and parameter uncertainties, and
(c) model uncertainties.
The major sources of the first type uncertainties are discussed in Section 8.2. This is fol-
lowed by a discussion of the data and parameter uncertainties in Section 8.3 and model
uncertainties in Section 8.4. Uncertainties associated with the future evolution of the dis-
posal system and justification step of the assessment, introduced in Section 6.6.3, are dis-
cussed in Section 8.5. Various methods that can be used to reduce the second and third types
of uncertainties in safety assessments, including deterministic and probabilistic approaches
to reduce data and parameter uncertainties, are discussed in Section 8.6, and verification
and validation to reduce model uncertainties in Section 8.7.
8.2 UNCERTAINTY IN THE UNKNOWN FUTURE STATE OF THE DISPOSAL
SYSTEM
The major sources of uncertainty in the assessment of a radioactive waste disposal system
are without a doubt the future evolution of the system. One of the major difficulties experi-
enced in this regard is that a post-closure assessment must often be performed even before
the disposal system is in place. The assessment could therefore be influenced adversely by
factors such as potential differences between the original design of the repository and the
eventual facility, as well as the intended and actual waste emplacement schedule and con-
figuration. Factors such as these can, to a certain extent, be controlled on a regulatory basis.
However, there are others, such as the habits and behaviour of people in the future and
natural and human induced disruptive events and processes, which cannot be controlled in
this way, and must be addressed explicitly in the assessment.
The approach commonly followed today to address the future evolution of a disposal
system is scenario generation, introduced in Section 6.6.3. In scenario generation, per-
sonal judgement is used to assign individual probabilities to each of the postulated
scenarios that can be compared with a prescribed riskstandard, without the restriction
that the probabilities of the selected scenarios should sum to unity. This is not a trivial
task. A structured approach is therefore necessary to ensure that the process is system-
atic, well documented, and that the uncertainties have been addressed adequately to
arrive at a credible decision. To assign probabilities to each of the individual postulated
scenarios may not be mathematically rigorous, but it is more useful than other approaches
(Kozak et al., 1999). The key point here is to use all the available information, and
information that can be derived from the analysis itself, to justify the inclusion of a
specific scenario in the assessment.
Uncertainties in Safety Assessments 113
8.3 DATAAND PARAMETER UNCERTAINTIES
The term datum (plural data) is defined in the Oxford English Dictionary as: A thing given
or granted; something known or assumed as fact, and made the basis of reasoning or calcu-
lation; an assumption or premise from which inferences are drawn. However, the term will
be used here exclusively to denote directly measurable and not derived quantities, for which
the term parameter will be used. For.exam ple, the term data could refer to measurements of
water levels in boreholes, while the hydraulic conductivity and storativity are parameters
that can be derived from the measurements.
Uncertainties in the data can only arise from measurement errors, according to the previous
definition, for which there are essentially two sources: instrument and human errors. It is
usually not too difficult for an observer to limit the influence of instrument errors, which are
mainly caused by the imprecision and malfunctioning of the available measuring devices,
but human errors are often difficult to detect. A few of the major human sources include, for
example, incorrect or misapplied measuring techniques, systematic errors and blunders. A
common source of systematic errors in safety assessments is where measurements are taken
on disturbed samples, because in situ measurements are impossible.
Measurement errors will obviously carry over to parameter errors and therefore uncertain-
ties. However, there are additional sources of parameter uncertainties. One source often
encountered in safety assessments is the large spatial and temporal variation in many of the
data sets. It is therefore not always possible to obtain sufficient data, or ensure that the
methods used in deriving the parameters are above suspicion. A common approach to cir-
cumvent this, so-called insufficient data uncertainty is to say that the person responsible for
the analysis of the data should apply his or her personal judgement. However, nature be-
haves in very subtle ways. The approach can therefore add even more uncertainty to the
data, unless the analyst can demonstrate explicitly that his judgement is supported by the
observed behaviour of the phenomenon.
Another major source of parameter uncertainty is that the data used to derive a parameter
may not be representative of the parameter. Two of the most common effects responsible for
this are: scale and geometric effects. Scale effects can often be observed in situations where
parameters derived from laboratory measurements are used to describe a phenomenon in
the field. Geometric effects can arise on various occasions, but the one most commonly
observed in practice is associated with the geometry of the aquifer. Most of the mathemati-
cal models employed in the study of ground water flow, including those discussed in Section
6.2, are based on the conventional geometry of a porous medium. Considerable errors could
Uncertainties in Safety Assessments 114
therefore arise if such a model is applied blindly to an aquifer in fractured rocks, or the
multi-porous fractured aquifers of the Karoo formations in South Africa.
8.4 MODEL UNCERTAINTY
8.4.1 General
Although data and parameter uncertainties will certainly affect model uncertainty, there are
a number of additional uncertainties that are particularly important for the development of
models in the analysis of scenarios. These uncertainties arise in all the model types dis-
cussed in Section 6.2-mathematical, conceptual, analytical and numerical models. Nu-
merical and many of the analytical models are usually implemented on a computer through
computer codes that are often quite complex. It is consequently not always possible to
prove that the codes implement the conceptual models correctly and do not contain pro-
gramming errors. This adds an additional source of uncertainty referred to here as code
uncertainty. These uncertainties will now be described in more detail.
8.4.2 Mathematical Model Uncertainty
There are basically three sources of uncertainties associated with the formulation of a math-
ematical model for a physical phenomenon, according to its definition in Section 6.2.3:
(a) An incomplete knowledge of the features, events and processes associated with the
phenomena.
(b) The inability to express many of the features, events and processes in exact math-
ematical terms.
(c) Errors made in the development of the mathematical model.
Although these uncertainties can occur in all models used in safety assessments, they are
especially important for the phenomenological models introduced in Section 7.4. As men-
tioned there, the intention with these models is not to explain the behaviour of a phenom-
enon, but only to describe it. Since it is impossible to predict the behaviour of a phenom-
enon if one cannot explain it, a completely different phenomenological model (or coeffi-
cients, in the case of transfer coefficient models) may be needed to describe the phenom-
enon in the future. Such variations can cause large uncertainties in the computed potential
dose or the risk to human health in safety assessments.
It is well-known that many of the constitutive parameters used in deterministic models,
Uncertainties in Safety Assessments 115
such as the hydraulic conductivity and specific storativity of an aquifer, vary in space and
therefore subject to parameter and insufficient data uncertainty. What is quite often ne-
glected though is that the parameters are also intrinsic functions of time. Although this
dependence may be weak, there are indications that some of them can vary considerably
over periods of a few decades. This applies in particular to the hydraulic parameters of the
highly elastic Karoo aquifers of South Africa (Botha et al., 1998). Since this dependence
cannot be derived from the underlying theory, special attention need to be given to the time-
dependence of these parameters, either during the site characterization or pre-elesure as-
sessment of a radioactive waste disposal system.
8.4.3 Conceptual Model Uncertainty
The term conceptual model was defined in Section 6.2.4 as the governing equation, or set of
equations, with its associated constitutive parameters, boundary and initial (if required)
conditions, to describe a specific realization of a given phenomenon. This may not always
be an easy task. For example, it is very laborious and expensive exercise to determine the
specific storativity and the three-dimensional hydraulic conductivity tensor of an aquifer
(Hsieh et al., 1983). It is therefore often tempting to describe a given phenomenon with a
less sophisticated conceptual model. An approach that is frequently applied in such cases, is
to replace the three-dimensional mathematical model with a reduced two-dimensional math-
ematical model in the conceptual model for such phenomena (Botha, 1996). This approach
may be acceptable when the conceptual model will only be used for a short period (a few
decades), but can yield completely wrong results if the model is to be used for a hundred (in
some cases even less) years.
A similar situation arises in those cases where the conceptual model includes a differential
equation and one has to prescribe suitable boundary and initial conditions. Although it is
usually not too difficult to prescribe suitable initial conditions, the exact boundaries of many
phenomena, especially those related to the geosphere, are often not known. A common
practice in such situations is to use observations of a phenomenon, such as water levels, to
prescribe suitable boundaries and boundary conditions for the conceptual model. Many
boundaries used in conceptual models therefore do not reflect the actual extent of the phe-
nomena. The possibility therefore exists that an assessor may exclude regions from his or
her model that will be influenced by the phenomenon over the time scales relevant to a
safety assessment of radioactive waste disposal sites. The combination of choosing a wrong
conceptual model and wrong boundaries will certainly introduce considerable uncertainties
in the final results of the safety assessment of a waste disposal site--even lead to meaning-
less results. Special care should therefore be exercised in choosing appropriate conceptual
Uncertainties in Safety Assessments 116
models in safety assessments of waste disposal sites.
8.4.4 Numerical and Approximation Uncertainty
As mentioned in Section 6.2.5 the governing equations of a conceptual model must be
solved either analytically or numerically for the model to be useful. In practice, this means
that one must be able to assign specific numbers to the conceptual model. The equations
that appear in many of the conceptual models cannot be expressed in terms of simple
arithmetic expressions, such as polynomials, with the result that they must be first
approximated with simple arithmetic expressions. This will introduce a degree of what will
be called approximation uncertainty into the final result. Approximation uncertainties can
arise from a number of sources such as the truncation of an infinite series. However, the
largest source of approximation uncertainty is probably that associated with the discretization
of a domain when solving a differential equation through a numerical method such as the
finite difference or finite element method.
Approximation uncertainties can be limited considerably by studying the convergence prop-
erties of the approximation. Unfortunately, this is often considered as too time consuming
and expensive, and therefore simply neglected in many modelling exercises-a luxury that
a safety assessor cannot afford.
Computer arithmetic has advanced considerably over the past half century, to such an extent
that it is now possible to represent a number with thousands of digits if necessary (Wolfram,
1996). However, it is doubtful if there is enough computer resources in the world to carry
out the computations in safety assessments to this accuracy. It will therefore be unwise to
completely neglect numerical uncertainty in the safety assessment of radioactive waste dis-
posal systems.
8.4.5 Computer Code Uncertainty
Coding errors and user errors mainly cause uncertainties in computer codes. Coding
errors can vary from straightforward blunders to more subtle errors, such as typing I for
1, and the inefficient or deficient implementation of convergence and stability criteria.
However, the majority of models used in safety assessments have usually been tested
thoroughly for coding errors. User errors are therefore often the major sources of com-
puter code uncertainties, because of the complexities of the models and processes in-
volved in the safety assessment.
Uncertainties in Safety Assessments 117
8.5 TREATMENT OF FUTURE UNCERTAINTIES
8.5.1 General
It follows from the discussion of the framework for a post-closure safety assessment in
Section 6.5 that the formulatiori and the analysis of well-justified scenarios will form an
integral part of such an assessment. According to OECD/NEA (1998), Herman Kahn intro-
duced this method of analysing futuristic events in the 1950s. The method was initially used
to study the various options civil defence authorities could take in response to a thermal
nuclear war. The method has since become a standard tool in decision and statistical analy-
ses of future events.
The term scenario as defined in Section 6.6.3 and used in this thesis, is associated solely
with the anticipatedjUture evolution of the disposal system, as opposed to its normal evolu-
tion. The main purpose of scenario generation in the post-closure safety assessment of a
radioactive waste disposal system is thus essentially to use scientifically-informed expert
judgement to guide the development of descriptions of the disposal system and its future
behaviour. However, it does not try to predict the future; rather, the aim is to identify salient
changes, based on analysis of trends, within which variants are explored to investigate the
importance of particular sources of uncertainty. The emphasis is therefore on providing
meaningful illustrations to assist the decision process (Watts et al., 1999).
The first step in the generation of scenarios for such a system is to identify a comprehensive
list of Features, Events and Processes (FEP) associated with the components of the disposal
system, described in Chapter 4. Since not all the FEPs in the list will be equally important
for a specific site, this list is systematically screened and only those that are significant for
the underlying assessment context and system description are retained. The relationships
between these FEPs are then investigated in detail and used to establish a list of appropriate
scenarios for the site.
New approaches to scenario generation and the uncertainties associated with the future
state of a disposal system are developed continuously. Despite differences in approaches
and ordering of the steps, the concepts behind the steps are the same for all scenario genera-
tion procedures. Before a systematic approach to generate scenario are discussed, a sum-
mary will be presented of the work that has been done on the development of a comprehen-
sive FEP list, methods to screen the FEP list, and methods to order the screened FEPs in a
structured manner to generate scenarios.
Uncertainties in Safety Assessments 118
8.5.2 Developinga Comprehensive List of FEPs
The first step in scenario generation is to compile a list of all FEPs related to the specific
waste disposal site. Since the compilation of such a list is a creative, multi-disciplinary task,
that may depend exclusively on scientific reasoning and expert judgement (Bonano et al.,
1990), various attempts have been made to establish an independent, comprehensive list,
which can be used to establish site-specific lists. The latest of these lists is the one devel-
oped for geological disposal systems by the Working Group on Scenario Generation of the
OECD Nuclear Energy Agency (OECDINEA, 1998). This document also provides a good
summary of the experience gained in developing such a FEP list with its associated databases
and scenarios.
The Scenario Generation and Justification Working Group of ISAM (see Section 6.1) are
currently addressing a similar investigation of FEPs related to near-surface disposal sys-
tems. This Working Group adopted the International FEP list of the NEA Working Group
and expand it to include FEPs relevant to near-surface disposal systems (IAEA, 1998b).
The current version of the list, known as the ISAM International FEP List, is given in Ap-
pendix A. Other sources of information for these systems are the report of Sumerling et al.
(1989) on scenario generation for the Drigg facility in the UK, the preliminary safety analy-
sis for the Canadian Intrusion Resistant Underground Structure (AECL, 1997), and the
assessment of the Greater Confinement Disposal Facility at the Nevada Test Site (Guzowski
and Newman, 1993).
8.5.3 Screening of the FEP List
It is simply impossible to consider all FEPs in a comprehensive FEP list, from a practical,
technical and economical point of view. Fortunately, some of the FEPs may not be applica-
ble in terms of the assessment context, existing legislation, regulatory requirements, char-
acteristics of the site or disposal concept, or are unlikely to be present at a specific site. The
second step in scenario generation is thus to screen the list and discard irrelevant FEPs with
the necessary justification. The methods used in such a screening are usually based on (a)
expert judgement, (b) probability of occurrence and (c) consequences of the FEPs (Kozak
et al, 1999). In the so-called Sandia Scenario Approach, which stems from the work of
Cranwell et al. (1990), fault trees are used to represent key aspects of a scenario. The branches
of a tree are then assigned probability values that may be propagated to produce an overall
probability of occurrence of the scenario. In later versions of the approach (Guzowski,
1990), FEPs are screened based on expert judgement of either the likelihood that they may
occur or their consequences. The full scenario of remaining FEPs are then screened, based on
Uncertainties in Safety Assessments 119
either the propagated probabilities of the FEPs or expert judgements of their consequences.
The Sandia Approach will not be discussed further here, because of a number of disadvan-
tages such as: the tendency to generate a large number of scenarios that must be evaluated
simultaneously, and its poor ability to address time-dependent variables. The method was,
moreover, developed for particular regulations in the US high-level waste programme
(Chapman et al., 1995). The probabilities assigned by the method are also primarily derived
from expert judgement. Significant resources for elicitation of judgements is, therefore,
required when applying the method. There is thus a risk that the results will be biased by
these judgements and that they have more mathematical significance than is actually the
case (Kozak et al., 1999).
An alternative method for screening FEPs, which will be used in this thesis, was introduced
by Eng et al. (1994). This method differentiates between FEPs acting within the temporal
and spatial boundaries of what they call the Process System and those that act from external
the temporal and spatial boundaries onto the Process System-the so-called scenario-gen-
erating FEPs (EFEPs). The Process System itself is defined as the organized assembly of all
FEPs required for description of barrier performance and radionuclide behaviour in a
repository and its environment, and that can be predicted with at least some degree of deter-
minism from a given set of external conditions (Eng et al., 1994). This division corresponds
very much with the conceptual representation of a disposal system in Figure 4-1. The terms
Process System and disposal system as used in this thesis can therefore be considered as
synonyms. A scenario in the Process System approach is then defined by a specific set of
external conditions, which will influence the FEPs in the Process System (Eng et al., 1994).
8.5.4 Ordering of FEPs
The next step in scenario generation is to order the screened FEPs into a coherent structure
suitable for a consequence analysis. The increasing scrutiny of safety assessments analysis
by regulatory bodies, and the public, have led to increasing requirements for scenario gen-
eration and justification of the associated models. Three methods have been described in the
literature that can be used for the ordering of FEPs: lists and tables, influence diagrams and
the Rock Engineering System (RES), matrix approach.
The ordering of FEPs for near-surface disposal systems have usually been limited to lists
and tables (Sumerling et al., 1989; AECL, 1997), while the influence diagram and RES
matrix approaches have received considerable attention for geological disposal systems
(Eng et al., 1995; Skagius and Wiborgh 1994; Skagius et al., 1995; Smith et al., 1996;
Uncertainties in Safety Assessments 120
BIOMOVS IT, 1996, Kozak and Wei Zhou, 1998). Skagius and Wiborgh (1994) who evalu-
ated the use of influence diagrams concluded that the method is very useful in the develop-
ment of FEP interactions, but that it requires a considerable effort to develop the influence
diagrams. They also claim that influence diagrams do not significantly improve the trans-
parency of the FEP process, because it is difficult to use and read the diagrams. The RES
. matrix method, on the other hand, increases the transparency of the ordering considerably
(Kozak and Wei Zhou, 1998), and makes it possible to trace judgements made during model
development (Skagius et al., 1995). The method also provides a good basis for interaction
between regulators and site developers. This method will consequently be used in this thesis.
8.5.5 A Systematic Approach to Scenario Generation
The integrated approach to scenario generation presented in this thesis considers the ISAM
International FEPs list as the primary reference point to ensure that all issues relevant to the
assessment are covered. The subsequent screening and categorisation of such a list gener-
ates an audit trail, allowing the analyst and regulators to scrutinize the logic of the argu-
ments supporting the definition of models and the calculations effectively.
The first step in this approach is a clear definition of the extent, nature and content of the
Process System to be analysed; thereby enabling all potentially relevant FEPs to be catego-
rised either as Process System FEPs or EFEPs for the purpose of the assessment. The choice
of EFEPs or scenario-generating FEPs is based on expert judgement, but in a qualitative
sense (Kozak et al., 1999), thereby avoiding the difficulties associated with the fault tree
analysis of the Sandia Approach. These scenarios are also not intended to be mutually ex-
clusive or comprehensive, since there is no intent to apply probability theory to them. Fig-
ure 8-1 gives a broad outline of the Process System PEPs and EFEPs relevant to the disposal
system. As can be seen from the figure, the assessment context defines the framework for
the identification of all relevant FEPs, while the EPEPs represent more' global' phenomena,
which act as a boundary condition for the disposal system domain. Fundamental to the use
of the Process System is the identification of the boundary of the Process System, the im-
portant components on either side of that boundary and their interrelationships (Watts et al.,
1999).
The Process System itself and its associated FEPs defines the anticipated normal evolution
of the disposal system and its environment and provides a necessary central theme in report-
ing on the radiological performance ofthe disposal facility. This requires a conceptualisation
of the Process System, based on its component features and characteristics, and an appraisal
of potential pathways for the released waste to the environment. The Process System FEPs
Uncertainties in Safety Assessments 121
are therefore included implicitly in scenario generation, and does not represent a scenario in
itself. With this configuration, the afety performance of the Process System can be as-
sessed independently of the influence of EFEPs. This may be unrealistic as changes in the
environment, for example, are neglected. However, it constitutes a useful benchmark against
which the significance of other results can be compared and it represents a practical basis
for exploring sensitivities to parameters and modelling uncertainties Watts et al. (1999).
From a scenario-generation point of view, it is necessary to perform a structured screening
and review of all EFEPs, taking into account prescribed screening arguments and the over-
all assessment context. As an example, the screening argument currently anticipated for the
Drigg safety assessment include the following (Watts et al., 1999):
(a) physically implausible given the site context or time scales of concern,
(b) rate or probability small relative to other EFEPs,
o ASSESSMENT CONTEXT
I EXTERNAL FACTORS
1.1 Repository 1.2Geological 1.3Climatic 1.4 Future human 1.5Other
factors processes & events processes & events actions
2 DISPOSAL SYSTEMDOMAIN: ENVIRONMENT FACTORS
2.1Waste & 2.2 Geological 2.3 Surface 2.4 Human behaviour
engineered features environment environment
3 RADIONUCLIDE AND CONTAMINANT FACTORS
3.1Contaminant 3.2 Release and 3.3 Exposure factors
characteristics migration factors
IMPACT
Figure 8-1 Outline of the ISAM international list of features, events and processes rel-
evant for near-surface disposal systems (IAEA, 1998h).
Uncertainties in Safety Assessments 122
(c) global disaster,
(d) included elsewhere, and
(e) excluded by assessment context or regulatory guidance.
This screening will serve as a basis to distinguish those external factors that represent a
significant influence on the long-term evolution of the disposal system. The EFEPs are
therefore included explicitly in scenario generation.
The Process System FEPs and screened EFEPs are next used to develop and justify alterna-
tive conceptualisations of the Process System, including appraisals of the potential path-
ways for the released waste to the environment. For each of the conceptualisations, outline
scenario descriptions (alternative futures) are also derived from a consideration of the inter-
actions between the Process System FEPs and the EFEPs system. A structured representa-
tion of the interactions will provide an audit trail and facilitate conceptual model develop-
ment. As discussed in Section 8.5.4, this is best achieved using the RES matrix approach.
The basic device used in the RES matrix approach is the interaction matrix, in which the
main variables or parameters are identified and listed along the leading diagonal of a square
matrix. The interactions between the parameters occur in the off-diagonal terms. A simple
example of a 2x2 matrix is illustrated in Figure 8-2 with the concrete container and backfill
Parameter A Interaction
A to B
-- --- --- --- --- --- -~
Concrete
Container Leaching
Interaction Parameter B
BtoA
~- --- --- --- --- --- --
Change in
Chemistry Backfill
Figure 8-2 A simple 2x2 interaction matrix illustrating the application of the RES ma-
trix method to interactions between features, events and processes in a ra-
dioactive safety assessment.
Uncertainties in Safety Assessments 123
as diagonal elements. Leaching represents an interaction between the container and backfill,
which in turn will introduce a change in the chemistry of the backfill that could influence
the concrete container again. This defines a clockwise convention for the influence direc-
tion. The matrix can of course be easily extended to an nxn matrix, for a system with Il main
parameters.
In the Process System approach, the components of the disposal system can also be in-
cluded in the interaction matrix, and analysed in greater detail by creating one ore more
sub-matrices, as shown in Figure 8-3. This suggests that the elements on the main diagonal
can be represented by a specific theme, such as the migration pathway of radionuclides
from the waste to humans. The off-diagonal elements represent the interactions of FEPs
Interaction
Entity I
Entity I Bcl~kfill and
.: Entity 4
I---t---t---r----,l"""l;eosphere---+--t-----t---'
1__ I------l__ r~~~:~~~-I---t-I--fthaeirOeSI'
Figure 8-3 A conceptual interaction and secondary interaction matrix for the Process
System approach with the components of the disposal system included ex-
plicitly.
Uncertainties in Safety Assessments 124
that cause or influence the migration of the radionuclides from one diagonal element to
another along the identified pathway. Those above the diagonal represent the influence on
forward movement, while those below influence the backward movement. This is illus-
trated in Figure 8-4, which represent a 5x5 matrix and the potential migration pathway of
radionuclides from element 0, through various interactions between diagonal and off-di-
agonal elements, to element E.
It is important to note that there is no rigorous way of using the RES matrix. It is merely an
auditing tool to order FEPs in a structured and traceable manner to facilitate scenario gen-
eration and conceptual model development.
Figure 8-4 Principle of a migration pathway of radionuclides through the interaction
matrix.
8.5.6 Model Development for Consequence Analysis
By following the migration pathway of radionuclides in Figure 8-4, one can identify all the
processes, interactions and phenomena that may influence the migration and must be ac-
counted for with heuristic or phenomenological mathematical model . Some of the proc-
esses can be coupled and included in one mathematical model, such as the moisture move-
ment through the concrete barriers, back-fill of the near-field, and the unsaturated zone.
However, the degradation of the concrete structure will require a different mathematical
model.
Uncertainties in Safety Assessments 125
Although it will probably be rarely necessary, new mathematical models have sometimes to
be developed for features that could influence the decision adversely, but is not described
adequately by existing models, or if no model exists for the feature.
Once all mathematical models needed to represent the migration of radionuclides along a
chosen pathway are identified, the information from the conceptualisation of the Process
System can be used to develop a conceptual model for the migration. This include, for
example, the domain over which the mathematical model is to be applied, initial and bound-
ary conditions and any constitutive parameters required to solve the mathematical model.
This defines the conceptual model for the scenario.
8.6 TREATMENT OF DATAAND PARAMETER UNCERTAINTY
Uncertainties in the quality and accuracy of data and parameters can often be considerably
reduced in safety assessments, by applying suitable quality assurance procedures during the
collection and manipulation of the data. However, many of the data required in a post-
closure safety assessment are not known, or will ever be known, by the assessor. It is there-
fore impossible to eliminate data and parameter uncertainties completely from safety as-
sessments.
There are essentially two types of methods that can be used to study the influence of data
and parameter uncertainties in safety assessments-deterministic methods and probabilistic
methods. Indeterministic methods a set of estimated values of a parameter is used to evalu-
ate the impact that the parameter may have on the final results, particularly the regulatory
decision.
Probabilistic methods, such as the Monte Carlo method, represent a parameter through a
probability density function (pdf), or the mean and variance of such a distribution. These
values are then used in a series of analyses to arrive at a probability that the regulatory
decision will be achieved or not. It stands to reason that these pdfs must be carefully con-
structed if the uncertainty estimates for the model predictions are to be meaningful. In
classical statistics the pdfs are usually derived from known values of the parameter or data.
In practice this usually means that a large set of parameter or data values must be known,
before one can obtain estimates of the mean and variance of the data set and confirm the pdf
by the testing it with various hypotheses. A common approach in classical statistics is there-
fore to use whatever data is available at a specific time and recalculate the pdf whenever addi-
tional data becomes available, thereby increasing confidence in the pdf (Freeze et al., 1990).
Such an approach may not always be acceptable in the safety assessment of a waste disposal.
Uncertainties in Safety Assessments 126
The pdfs used in safety assessments are consequently often based on the Bayesian method.
To describe this method, let A and B be two sets of random variables each with its own so-
called prior probability functions, say peA) and PCB). Inmultivariate analysis it is sometimes
necessary to know what the distribution of A will be given, a set of observations on B. Let
P(AIB) denote this so-called conditional or posterior probability function of A. By using
Bayes's theorem (Groebner and Shannon, 1989), this posterior probability function can be
expressed, in the case where A and B are discrete variables, as
P(A.I B)
I
where it is assumed that both peA) and P(BIA) are known. The importance of this theorem
is that it allows one to derive a probability distribution for the variable A, even when very
little information is available on A, provided that P(BIA) is known. It is therefore sometimes
possible to derive a pdf for a given variable, even before any measurements are taken of the
variable through Bayes's theorem (Freeze et al., 1990). Such a pdf can consequently differ
considerably from the classical estimate when the data are sparse.
Confidence in a pdf derived with the Bayesian method can of course also be increased by
recalculating it every time more data become available, relying less on the conditional
information. The pdf should under these conditions converge towards the classical estimate.
However, there may not always be enough time to establish that the pdf will represent the
behaviour of the parameter far into the future. Nevertheless, there are situations where the
conditional information may be sufficient to derive a pdf with confidence. This is, for example,
the case of groundwater levels in open boreholes that tend to follow the topography of an area
quite closely, as long as the aquifer is not pumped excessively (Van Sandwyk et al., 1992).
The deterministic and probabilistic methods each have certain advantages and disadvan-
tages, as far as the implementation of safety assessment methodologies is concerned. The
biggest advantage of the deterministic approach is its simplicity of implementation and
transparency of interpretation. However, deterministic variations in the parameters will not
necessarily yield bounds on the final result. The approach therefore does not lend itself to
the development of a clear decision criterion that can be used to determine whether the
facility complies with the regulatory decision or not (Kozak, 1997).
Although the probabilistic method can provide a criterion for the regulatory decision, the
criterion need not necessarily yield a statistically rigorous estimate of site's performance,
Uncertainties in Safety Assessments 127
because of the uncertainties involved in deriving the necessary pdfs. All that the method
really can do is to provide the analyst with quantitative information about the relative like-
lihood of various estimates of the performance. Kozak et al. (1993) suggests that this inde-
terminacy can be reduced in the case of the Monte Carlo method, by coupling the method
with a stratified parameter sampling strategy, such as the Latin Hypercube Sampling method
(Iman and Shortencarier, 1984). Since it is already difficult to interpret probabilistic meth-
ods and to communicate the results to non-expert audiences (Kozak 1997), this extension
may not always be acceptable. Another argument often raised against probabilistic methods
is that they require a considerable computational effort and resources, although this argu-
ment is loosing ground with the rapid increase in computing power in recent years.
In summary it can be said that neither the deterministic nor the probabilistic method has
special advantages as far as the reduction of parameter uncertainty in the safety assessment
of radioactive waste disposal sites is concerned. Amore pragmatic approach-use the method
that suites a particular site or pathway the best-would therefore seems to be more practical.
8.7 TREATMENT OF MODEL UNCERTAINTY
There do not exist specific guidelines to study model uncertainties, similar to those for
parameter and data uncertainties, except for the verification and validation of the concep-
tual model. Otherwise one has mainly to rely on strict quality control over the calculations
and the application of various assurance procedures.
The first uncertainty that may arise in the application of models is to what extent does the
mathematical model, on which the conceptual model for the phenomenon to be modelled is
based, represent the physical processes responsible for the phenomenon. The mathematical
models normally used in safety assessments have usually been tested thoroughly for this.
One can therefore safely neglect mathematical model uncertainties in safety assessments,
when comparing it with the other uncertainties in the assessment.
The main contribution to model uncertainty comes from the ability of the chosen concep-
tual model to represent the phenomenon. As discussed in Section 6.2.4, the conceptual
model represents the assumptions made to describe a specific realization of a natural phe-
nomenon. These assumptions represent, amongst others, the domain of interest, the
dimensionality of the problem and the initial and boundary conditions that may be needed
to solve any differential equations that appear in the conceptual model. Very little work has
been done to treat conceptual model uncertainty (Kozak et al., 1993). One approach is to
use a set of conceptual models that represents different types of boundary conditions, the
Uncertainties in Safety Assessments 128
spatial and temporal extent of the domain, may be even different processes, and compare
that with site-specific data. This approach is commonly referred to as the verification and
validation of the conceptual model.
Model verification can be described as the process of showing that the chosen conceptual
model and the analytical or numerical method used to solve it behave as intended. In other
words, the model must be able to reproduce observations of the phenomenon for which it
was developed. Model validation, on the other hand, is a process to show that the model is
able to describe observations of the phenomenon not used in its verification. Unfortunately,
it is not always easy to determine what component of the conceptual model is incompatible
with the phenomenon if the process fails.
Uncertainties associated with the numerical methods, used in the implementation of the
conceptual model, can be studied by comparing the results with an analytical solution of a
similar model, or by studying the convergence properties of the model.
CHAPTER9
A DECISION ANALYSIS FRAMEWORK FOR POST-CLOSURE
SAFETY ASSESSMENTS
9.1 INTRODUCTION
The post-closure safety assessment of a radioactive waste repository was described in Chapter
6 as a decision tool to determine the conditions under which compliance with safety objec-
tives can be reasonably assured. Such a safety assessment therefore very much resembles
system analysis, as developed in the operational research literature (Kozak, 1994), at least
in principle. However, as shown by the discussion in Chapters 7 and 8, a post-closure safety
assessment differs in one important aspect from a classical system analysis-the members
of the system are subject to large uncertainties. No attempt has therefore been made to date
to formally include system analysis theory into the post-closure safety assessment of a ra-
dioactive waste disposal site, nor is it the purpose of this thesis to do so. What will be done
here is to present a decision analysis framework that will aid in making feasible decisions in
such a safety assessment.
The decision analysis approach that will be followed here is based on a revision of the
hydrogeological decision analysis framework proposed by Freeze et al. (1990), which is
particularly suited for systems with large uncertainties and high risks. The basic philosophy
behind the methodology is that the risk that an engineering design will fail to meet design
objectives depends on uncertainties in the technical analysis. However, decision-makers
often base their decisions more on economic than technical factors in such analyses. It
therefore makes sense to also introduce the economic factor into the decision framework.
Indeed, this was one of the reasons why Freeze et al. (1990), introduced their decision
analysis framework.
The importance of the economic factor from an integrated radioactive waste management
perspective was introduced in Section 3.6.3. As used there, nuclear liabilities management
refers to both the reduction and discharge of nuclear liabilities. It is impossible to include a
full description of the framework here. However, the interested reader can find full details
in the references cited in Section 9.2, where the development and application of the meth-
odology is reviewed briefly. This is followed by a discussion of the basic principles of
decision analysis in Section 9.3. The framework proposed for the post-closure assessment
-129-
A Decision Analysis Framework for Post-closure Safety Assessments 130
of a radioactive waste repository is discussed in Section 9.4.
9.2 REVIEW OF THE DECISION ANALYSIS FRAMEWORK
Massmann and Freeze (1987a, 1987b) introduced the basic philosophy behind a decision
analysis framework in two papers on hydro geological projects. The first paper describes the
methodology for the risk-cost-benefit analysis in the design of new waste management
facilities. The second paper applies the methodology to assess alternative design strategies
for such a facility, and how a regulatory agency can use the analysis to assess alternative
regulatory policies. This was followed by a comprehensive series of four papers that give
excellent discussions of risk-cost-benefit analysis. These pioneering papers laid the founda-
tion for the application of a decision analysis framework for hydrogeological projects.
The series of papers begin with a detailed description of the framework as a tool in
hydrogeology (Freeze et al., 1990) and the application of the framework to two case studies
in groundwater contamination (Massmann et al., 1991). The first case study involves the
selection of an optimal pumping rate for an extraction well to capture an existing contami-
nation plume, whereas the second evaluates the design of a leachate collection system for a
soil remediation facility. In the third paper (Sperling et al., 1992), the framework was used
to demonstrate its usefulness in solving geo-technical problems associated with the design
of a groundwater control system at an open pit mine. The fourth paper (Freeze et al., 1992)
provides an introduction to the concepts of data worth analysis in reducing the uncertainty,
and the use of Bayesian statistics to provide a criterion when to stop the collection of data in
iterative site investigation programs. One of the key advantages of the methodology is its
suitability to aid in the design of site investigation programs and monitoring networks and
to assess the potential financial worth of additional data.
There are essentially two questions that are very important in many hydro geological inves-
tigations that can be answered by such a decision analysis (Freeze et al., 1992). The first is:
"Given a set of data on the hydrogeological parameters, which design alternative
from an available suite of design alternatives is the best one?"
and the second
"Is it worthwhile trying to improve the data set by taking additional measurements,
before deciding which is the best design alternative?"
The decision analysis framework has been applied to a large number of hydrogeological
investigations since the publication of the four papers. For example, James et al. (1996a)
A Decision Analysis Framework for Post-closure Safety Assessments 131
use the framework to determine what is the best remedial action for a contaminated aquifer,
what are the most important parameters to sample in such an investigation, and to what
extent will the collection of additional data be cost effective. James et al. (1996b) further
show that such an analysis can be a useful tool to help site managers to effectively allocate
the limited remediation funds and resources in cleaning the large number of contaminated
sites throughout the United States.
Jardine et al. (1996) apply the method to the design of a monitoring network at a waste
management facility, overlying fractured bedrock, with the objective to detect contaminants
before they reach a regulatory compliance boundary. The same approach was also used by
Bugai et al. (1996) to evaluate alternative management strategies for the water wells at
Pripyat Town, where the groundwater is contaminated with 90Srfrom the Chernobyl accident.
Janse van Rensburg (1992) introduced an interesting variation of the framework in his analysis
of the design of a well field, for the supply of water to a diamond mine in the Northern
Province of South Africa. His analysis not only takes uncertainties in the hydrological pa-
rameters into account, but also uncertainties associated with the financing of the scheme.
9.3 BASIC PRINCIPLES OF A DECISION ANALYSIS
The one essential element of a decision, as defined by Kirkwood (1997) is the existence of
alternatives. That is, one must have at least two different options of which only one can be
selected. There is a problem if only one option exists, but this is not a decision problem.
Most significant decisions involve situations where the various options lead to different
consequences or outcomes.
The models in safety assessment methodologies usually use only localised information to
describe the response of the environment to a specific option. They may therefore not be
able to identify an optimal management strategy if the same objective can be achieved with
more that one option. This can only be achieved through a thorough study of all alternatives,
requirements and constraints of the management strategy. This is the analysis referred to as
system analysis above. The analysis, which requires a clearly defined objective, can be
carried out in either an optimisation framework or a decision analysis framework.
In the optimisation framework the main aim of the analysis is to determine optimal values
for a given set of decision variables in a system, subject to a specified objective function and
a set of constraints. It is a general approach that will determine the best alternative from a
given set of alternatives, subject to the given objective function. Unfortunately, it is not
A Decision Analysis Framework for Post-closure Safety Assessments 132
always easy to determine the global extreme of a non-linear function. The approach is con-
sequently mainly restricted to problems with a linear objective function and constraints, in
practice. The approach, moreover, does not lend itself readily to a statistical interpretation,
with the result that it is not suitable for the analysis of risks or statistical variations in the
data, constraints and objectives (Freeze et al., 1990).
Decision analysis, described by Massmann and Freeze (1987a) as a formalisation of com-
mon sense for decision problems that are too complex to solve with informal common
sense, does not suffer from such limitations. Although the approach is less general than
optimisation, and therefore cannot provide an optimal solution across all the system vari-
ables, it is quite useful in eliminating alternatives that do not satisfy prescribed constraints
from a system.
9.4 DECISION ANALYSIS FRAMEWORKS
9.4.1 The Framework of Freeze
The original decision framework of Freeze et al. (1990) has six components, shown in the
. blue boxes of Figure 9-1,to which Janse van Rensburg (1992) added the hydrological and
financial components, shown in the orange boxes.
The purpose of the parameter, financial and geological uncertainty models is to study
uncertainties in these entities, as revealed by data obtained in the Field and Data Acquisition
Program. The uncertainties in the hydrological and hydrogeological parameters will usually
be based on actual field observations, while the financial uncertainties will depend on
variations in the financial markets. The analyses of these uncertainties should not present
any profound difficulties in practice per se, since there exists a number of well-known
models that can be used for this purpose (Freeze et al., 1990; Janse van Rensburg, 1992). A
similar situation also exists in the case of the engineering reliability model, where the
properties of most engineering materials are well-known from laboratory studies and practical
experience. The same can, unfortunately, not be said of the geological uncertainties.
As used here, the term geological uncertainties refers to all quantities related to what may
be termed the subsurface geometry. This includes, for example quantities such as,
discontinuities in the geology, the presence (absence) of a fluid conductive fracture networks
and planes, variations in the lithology and the tectonic and structural stability of the site.
These quantities are usually neglected in the majority of hydrogeological investigations,
with the result that their influence on subsurface flow is not well documented in the literature.
A Decision Analysis Framework for Post-closure Safety Assessments 133
Field Investigation
and
Data Acquisition Program
Hydrological Geological Hydrogeological Engineering
Parameter Uncertainty Parameter Reliability
Uncertainty Model Model Uncertainty Model Model
eological Financial
lation Simulation
ode I Model
Decision Model
Figure 9-1 Framework for hydrogeological decision analysis as proposed by Freeze et
al. (1990) and modified by Janse van Rensburg (1992).
The establishment of a suitable geological uncertainty model may, therefore, present the
analyst with his or her biggest challenge in developing a decision analysis framework,
particularly one devised for the post-closure assessment of a radioactive waste disposal site.
The main purpose of the various simulation models in the framework is to provide an estimate
of the behaviour of the particular component at a time t, so that the analyst can decide
whether the system still behaves within the prescribed constraints. The hydrological and
financial simulations are usually based on an existing stochastic model in these fields, while
the geohydrological simulations are based on either a numerical or analytical solution of the
flow and mass transport equations in Chapter 6. However, the model must then be applied
stochastically and not deterministically (Janse van Rensburg 1992), as is usually the case in
groundwater modelling.
9.4.2 The Disposal System Framework
A comparison of the framework in Figure 9-1 and the discussion of the uncertainties
associated with the post-closure safety asses ment of a radioactive waste disposal site as
discussed in Chapter 8, shows that the framework contains many of the uncertainties. This
suggests that it may be advantageous to use a similar framework in the post-closure assessment
of a radioactive waste disposal site.
A Decision Analysis Framework for Post-closure Safety Assessments 134
The ideal approach to design such a framework would have been to use existing data on the
post-closure safety assessment of a radioactive waste disposal site. Unfortunately, the idea
of a post-closure safety assessment has only been introduced very recently, as can be seen
from the discussion in Chapter 6. Indeed, the methodology itself is still in the development
phase. There is therefore not enough data (at least in South Africa) to use in designing such
a framework. However, a detailed review of the discussion in Chapters 7 and 8 suggest that
the one pre ented in Figure 9-2 may be quite suitable for this purpose.
As can be seen from Figure 9-2 all the components of the framework in Figure 9-1 are also
present in the new framework, albeit with different or additional functions. The engineering
reliability model, for example, now combines with the near-field uncertainties to determine
uncertainties in the source term, while hydrological uncertainties now affect the source
term, geosphere and biosphere. The geosphere uncertainty and simulation models will prob-
ably be similar to the geohydrological uncertainty and simulation models in Figure 9-1,
except that the geosphere models will always have to include mass transport.
The new components introduced in Figure 9-2 include the near field uncertainty and simu-
System Description
and
DataAcquisition
Engineering Near Field Geosphere
Reliability Uncertainty Uncertainty
Model Model Model
Near Field Hydrology
Simulation Simulation Simulation
Model Model Model
Biosphere Biosphere
Simulation Uncertainty
Model Model
Decision
Model
Figure 9-2 A possible decision analysis framework for the post-closure safety assess-
ment of a radioactive waste disposal site.
A Decision Analysis Framework for Post-closure Safety Assessments 135
lation models, the source term simulation model, and the biosphere uncertainty and simula-
tion models. The near field can be handled with models very similar to that used for the
geosphere, adjusted appropriately, but special models will have to be used for the source
term and biosphere models.
An important prerequisite for the application of a decision analysis framework is that the
various uncertainties must be known with a sufficient accuracy, so as not to skew the final
results. The common approach in classical statistics is to assume that the only pertinent
information about the uncertainty of a given sample of data is contained in the sample.
However, this is rarely the case in the application of a decision analysis framework where
the number of elements in the sample is often very small. Moreover, decision-makers feel
that the sample is not the only source of information on the data, and that subjective forms
of information, such as an expert opinion or personal experience, should be included in the
decision process. As pointed out in Section 8.6, this can be achieved by using Bayesian
statistics instead of the classical statistics. A decision analysis based on Bayesian statistics
is known as Bayesian decision analysis (Groebner and Shannon, 1989). A good summary of
previous applications of Bayesian decision analysis in geohydrology and hydrology can be
found in James and Gorelick (1994).
9.4.3 The Decision Model
The basic aim of the decision model is to give the analyst an indication whether a given
system will satisfy the objectives of a person who wants to implement the system. It usually
consists of a scalar function that measures the relative effectiveness of different alternatives
in achieving the specified objectives of the system. It is therefore comparable with the ob-
jectives of a post-closure safety assessment, discussed in Chapter 6.
It follows from the previous brief description that the decision model selected for a specific
study will very much depend OIl the purpose of the study and whether it should be applied in
an optimization sense or a decision sense. The type of constraint(s) placed on the system
will also be very important. For example, the operator of a water supply scheme may decide
to change some of the constraints, even though it may affect his objective adversely--an
approach that will not be possible if the objective is prescribed by a regulatory body.
A common approach used in the case where the system consists of a facility devised for
economic production purposes is to use an economic objective. In other words the facility
must be technically designed and operated in such a way that it maximizes the profit (or
minimizes the loss) of the owner. Let j be a member of sequence, J, of processes from
------------------------------------------------------------
A Decision Analysis Framework for Post-closure Safety Assessments 136
which the owner or operator of the facility may benefit. Let B.(t), C.(t) and R(t) denote this
) ) )
benefit and its associated cost and risk, respectively, at a time t during the lifespan, or other
time horizon, of the facility. The objective function for j can then be taken as the net present
value (NPV) of j taken over the chosen time horizon, discounted at the market interest rate
(Crouch and Wilson, 1982), and expressed mathematically as
T 1
<I>. = " (1+ I)' [B· (t) - C·(t) - R· (t)]
(9.1)
J ~ J J J
where Tis the time horizon (typically expressed in years) and 1 the market discount rate (a
decimal fraction). Notice that the three parameters in the square brackets, on the right hand
side of Equation (9.1) have to be expressed in monetary values. If the decision model is
applied to a radioactive waste management problem, then it is clear from Section 3.6.2 that
C(t) refers to the cost associated with the management strategy and by implication, contrib-
J
ute to nuclear liabilities. The risk, R.(t), in Equation (9.1) is often related to the expected
J
cost, C! (t), associated with the probability, P! (r), that the j-th member may fail in the year
J J
t, and expressed as :
(9.2)
where y. is a suitable weighting factor, introduced to account for the so-called risk-averseness
J
of many decision-makers. Owners and operators of facilities which do not have a strong
financial backing, may decide to choose y. > 1, while large companies more likely will take
J
a neutral approach and choose y. = 1. Some owners and operators may even want to use a
J
value of Y, that depends on C! (t). Notice that all the terms in Equations (9.1) and (9.2) are
J J
derived from economical considerations, except for pI, which will depend on the actual
J
behaviour of the j-th process.
It follows from the definition of the risk term in Equation (9.2) that this term will only occur
if the process j fails. However, notice that the term failure is used here to denote a situation
where the process fails to comply with certain criteria laid down for the decision analysis,
and not in its more common sense. For example, a failure may occur in a groundwater-
monitoring network if the concentration of a pollutant rises above an expected or regulatory
limit in one of the observation boreholes, the position of which may also be subject to some
form of control.
The cost, C! (t), above will only arise in the event that the process j fails. This may typically
J
include fines, taxes, and charges levied by a regulatory agency for failure to meet regulatory
standards, the cost of litigation should any arise, and the revenue foregone if the operations
A Decision Analysis Framework for Post-closure Safety Assessments 137
must be curtailed or stopped, because of the failure. The cost incurred in any remedial
action that may be necessary after such a failure should also be included in Cf (t) From a
J
radioactive waste management perspective, it is clear that these costs amount to another
source of nuclear liabilities that needs to be accounted for, as described in Section 3.6.2. It
is therefore often important to know exactly what caused a failure. In the case of a technical
facility this can often be achieved by studying a model of the process. This is the reason for
the inclusion of the uncertainty and simulation models in the decision analysis frameworks,
presented in Figure 9-1 and Figure 9-2.
The time or planning horizon defines the period over which the analysis has to be per-
formed, in other words, over which time period the system poses a risk to future genera-
tions. The pertinent discount rate is the market interest rate on borrowed money.
It is not difficult to show that the NPV value of future money approaches zero not very far
into the future (approximately 50 years in the case of a discount rate of 10%). The discount
rate and the time horizon are therefore not really independent in an NPV analysis. This has
obviously serious implications for the application of decision analysis in the post-closure
safety assessment of a radioactive waste disposal site, where the time frame of interest may
involve hundreds to thousands, even millions, of years. In this case it makes more sense to
compound the NPV at the inflation rate, rather than discount it at the interest rate. This
yields an objective function of the form (Gitman, 1988)
(9.3)
where I now represents the inflation rate.
To conclude this discussion of the decision model, it must be pointed out that the processes
or decision variables, as they are also called, in a decision analysis will vary from one
analysis to another. For example, in siting a new waste disposal facility where there is more
than one site to choose from, the sites themselves will be selected as decision variables,
while different designs will be the obvious candidates in designing a disposal facility for the
containment of waste. In designing a monitoring scheme, the monitoring positions, sampling
protocol and sampling frequency will all be suitable candidates for decision variables, or in a
remediation scheme the position and discharge rates of the pumping and injection wells.
There is little doubt that a decision analysis approach can make a significant contribution to
the management of nuclear liabilities, in particular to the reduction of nuclear liabilities, if
applied correctly. This statement is perhaps best illustrated through the following hypo-
thetical example.
A Decision Analysis Framework for Post-closure Safety Assessments 138
During the planning of a new radioactive waste repository for a specific type of waste, a
liability study with Equation (9.1) or (9.3), based on economic constraints alone, has shown
that a certain type of container and certain construction materials will yield the minimum
liability. However, a post-closure assessment of the repository's near-field revealed that the
interaction between the construction materials of the repository and the immediate geosphere
may create an environment that will corrode the selected container faster than its design
specifications. The possibility therefore exists that the repository could have been constructed
and even commissioned, if these two studies were performed separately and not related to
each other, with disastrous consequences.
9.4.4 Data Worth Analysis and Nuclear Liabilities
If the flow diagram of the safety assessment approach shown in Figure 6-1 are followed for
at least one iteration, the safety analyst will come to a point where he or she have to decide
whether the safety case was adequately demonstrated or not. In the latter case, a further
iteration of the safety assessment may be needed, which means that more information must
be generated, according to Figure 6-1. Such an exercise will clearly add to the nuclear
liabilities, even if the investigation is restricted to those parameters and assumptions that
have the greatest impact on the system performance. However, there is no assurance that the
additional data will reduce the uncertainty in the dose to acceptable levels. It would there-
fore be advantageous if the safety analyst can determine before hand if the additional data
will reduce the uncertainty in the dose sufficiently to justify the additional cost. One ap-
proach to solve the previous problem is to use a so-called data worth analysis as described
in Freeze et al. (1992), James and Freeze (1993), James et al. (1996a) and James et al.
(1996b).
There are two ways in which the worth of additional information can be estimated. The first
is to derive an expected value that perfect information will have on the safety assessment-
the so-called EVPI approach. This approach gives an estimate of the maximum impovement
that could be expected if the site conditions were known. The approach is consequently
usually used in conjunction with the second appoach, known as the expected value of sam-
ple information (EVSI) (James and Gorelick, 1994). This approach is based on the assump-
tion that the uncertainty in a sample of numbers decreases as the sample increases.
CHAPTERI0
CONCLUSIONS AND RECOMMENDATiONS
10.1 INTRODUCTION
The main objective of radioactive waste management and its underlying principles is to
ensure that human health and the environment are protected now and in future, without
imposing an undue burden on future generations (IAEA, 1995a). This implies that the im-
pactthe disposed waste will have on individuals and their environment must be determined
as a function of time, before any long-term management strategy of the disposed radioac-
tive waste can be implemented. This procedure is generally referred to as a post-closure
safety assessment. This term is a relatively new concept in waste disposal, introduced by the
IAEA around 1981 (IAEA, 1981) in an attempt to provide guidelines for the solution of one
of the most vexing problems in the world-the disposal of radioactive waste.
SouthAfrica produces radioactive waste through the full nuclear fuel cycle and the applica-
tion of radioactive materials in industry, research and medicine. There are two sites where
the waste from these activities are disposed at the moment-Vaalputs in the Northern Cape
Province and Thabana at Pelindaba near Pretoria. Vaalputs was selected after a detailed site
selection and evaluation program, but very little attention was paid to the post-closure safety
of the facility. Although it is known that Thabana was selected for 'the permanent storage of
radioactive waste', there is no indication that the site was ever subjected to a detailed site
selection or evaluation program. Indeed, all indications are that the site was merely selected
because of its proximity to Pelindaba, where most of the South African research activities
related to the nuclear industry were concentrated shortly after 1960. The result is that huge
uncertainties exist about the inventory of the waste disposed at the site and its future impact
on people and the environment.
One reason for the present situation in South Africa is that not much thought was given to
the impact that the disposal of radioactive waste may have on future generations. However,
there is a growing need to find a long-term solution for the disposal of spent fuel produced
by the AEC and ESKOM, in a manner that will ensure the safety of the public, and in
accordance with modem, internationally accepted, safety practices. Although some of the
other African countries are involved in activities related to the nuclear fuel cycle (e.g. neu-
tron activation or TRIGA type reactors), many of them posess spent nuclear sources. This
-139-
Conclusions And Recommendations 140
thesis describes a structured methodology that should contribute significantly to the solu-
tion of both problems, if implemented judiciously and conscientiously, and lay a sound
foundation for the long-term disposal of radioactive waste in the countries.
The proposed methodology-an integrated and structured approach to the post-closure as-
sessment of a radioactive disposal system-is to:
(a) ensure the safety of the present public and future generations,
(b) enhance the public acceptance of the methodology,
(c) keep the expenditure associated with the implementation of the methodology at a
rrummum,
The methodology recognises the interdependence between operational phase activities and
the post-closure behaviour of the disposal system. It is an iterative process that considers
site-specific, prospective evaluations of the post-closure phase.
The implementation of a post-closure safety assessment is a highly technical exercise, but
also contains significant social and political elements. Many of the elements, especially the
technical elements, will not be known at the beginning of the assessment. Moreover, it is
highly unlikely that a disposal concept can be introduced without the active participation of
the population. It is therefore of the utmost importance that the assessment should be per-
formed in conjunction with a research and development program aimed at reducing the
unknowns and promoting public awareness of the assessment.
Safety assessments of radioactive waste disposal sites have historically often been inter-
preted as an attempt to predict the future transport of radionuclides through the geosphere
and biosphere of the site, until ingested by a person. Many countries have consequently
built and developed surface and subsurface research laboratories to investigate this path in
detail (Kickmaier and Mckinley, 1996). However, the view expressed in this thesis is that it
will be difficult (if not impossible) to achieve such an objective. The proposed methodology
is therefore not aimed at satisfying the historical objective, but rather to convince all inter-
ested parties that the disposal system will satisfy all regulatory and other constraints within
reasonable limits. This objective can be achieved with considerable less effort and a corre-
sponding reduction in research and development costs and time.
10.2 CONCLUSIONS
The development of any, and not only radioactive waste disposal, systems will always cost
money. Some of the costs will be clearly identifiable, such as those associated with the
Conclusions And Recommendations 141
design and development of the site, installation of the necessary infrastructure and the op-
erational costs, but other are hidden. This is for example the case with liabilities that may
arise should the system fail in future, thereby causing damages to other people's properties
and lives. Although the origins of the two types of costs differ, they are still costs that have
to be met at some point in time. Moreover, these costs are not necessarily independent from
each other. For example, the failure of the system could probably have been prevented if
more attention was paid to its design and construction. The view taken here is therefore not
to consider the costs separately, but simultaneously in what is known as nuclear liabilities.
This has the advantage that one can use a decision analysis framework to manage all the
costs effectively. Since the costs associated with the planning and development of a radio-
active waste disposal system can be huge, it is essential that no post-closure assessment of
such a system should be undertaken without a decision analysis framework for the nuclear
liabilities. The advantage of the approach is that it allows one to link today's actions with
tomorrow's consequences within the limits of the prescribed regulatory and possibly public
constraints.
The first step to take in a post-closure safety analysis is to identify the type and nature of the
waste that must be disposed and then to select a suitable site. For example, the geology at
some of the prospective sites may support the disposal of low and intermediate level radio-
active waste in a near-surface disposal system, but not the disposal of high-level waste in a
geological disposal system.
There will probably not be very many sites that are technically suitable and acceptable to
the public for disposal of radioactive waste. Nevertheless, one of the sites has to be selected
as the disposal site at some point in time. One approach to solve this problem is to carry out
a site selection, as described in Section 4.3.2 and use this information in the decision analy-
sis framework, discussed in Chapter 9, with the sites themselves as decision variables. Al-
though the costs of these investigations will add to the nuclear liabilities associated with the
future acquisition and development of the chosen site, this will be more than accounted for
in the later stages of the post-closure assessment analysis. Moreover, a further analysis of
the results may already indicate that there is a possibility the site mayfail in the future and
to what extent the failure may add to the nuclear liabilities of the site. In such cases, includ-
ing an estimate of these liabilities in the decision analysis framework can further enhance
the site selection procedure.
Although the geological properties will play a very prominent role in site selection, one
should never neglect the type of waste. For example, it is commonly accepted that clayey
formations are ideal for the near-surface disposal of low- and intermediate level wastes,
Conclusions And Recommendations 142
because of their ability to retard the motion of the nuclides. While this may be true for most
of the types of radioactive materials, this is certainly not true in the case of radioactive waste
that consists mainly of carbon compounds. It is known that carbon compounds can increase
the hydraulic conductivity of clays by a factor of 104 and more (Quigleyand Fernandez,
1989), thereby almost nullifying their retention properties. Since this property of carbon
compounds is well-known, it makes sense to include it already in the site selection. How-
ever, other compounds may show a similar behaviour, but on a lesser scale and may not be
detected before a comprehensive site characterization, described in Section 4.3.3, is com-
pleted. This suggests that the waste acceptance criteria, discussed in Section 3.3.3, be in-
cluded in the site characterization. This will allow one to determine the inventory and con-
centration limits of the waste that can be safely disposed in the system. An IAEA initiative
is currently underway to address this problem in existing and proposed near-surface dis-
posal facilities with the following objectives (lAEA, 1999).
(a) To test the feasibility of a proposed disposal concept as a means to manage the
waste.
(b) Provide some guidance (be it only in order of magnitude) on the likely impact that
the disposal system may have on people and the environment.
(c) Assist with the 'screening process' in the organization and development of a near-
surface disposal system.
The derived activity values can also be used as a starting point by countries and organisa-
tions that are developing plans for a near-surface disposal, but do not have site-specific
activity values.
The next step after a potential site has been selected is to design and characterize a disposal
concept that will comply with the regulatory and other constraints. The information on the
site's geosphere and biosphere acquired during the preceding phases will provide valuable
information for the design, particularly the information on features that may act as natural
barriers. However, other factors, such as the characteristics of the waste, and the design of
the waste packages, engineered barriers and repository will all influence the design of the
disposal concept. This will require vast amounts of information on the system's near-field,
geosphere and biosphere, even if the more liberal approach advanced in this thesis is fol-
lowed. However, this can be reduced if the safety analyst's attention is focussed on the primary
objective of the investigation-to show that the disposal system will satisfy all regulatory and
other constraints within reasonable limits. It is believed that the most efficient way to achieve
this objective is to follow the safety assessment approach outlined in Figure 6-1 and the decision
analysis framework discussed in Chapter 9. As shown in Figure 6-1 the analyst will reach a
Conclusions And Recommendations 143
point where he or she has to decide whether the safety case was adequately demonstrated or
not. In the latter case, the analyst will have to assess what additional information will be
needed. In many cases this will entail the acquisition of more detailed system-specific data,
especially on those parameters and assumptions that have the greatest impact on the sys-
tem's performance. However, as shown by the discussion of data worth analysis in Chapter
9, there is a cost-effective limit on obtaining the additional data. The limit is, however, not
fixed but varies with the type of data and the purpose of the investigation. It may therefore
be necessary to perform a decision analysis of the nuclear liabilities at this point to see if the
additional costs can be justified, before proceeding with the acquisition of the data.
An important feature of the proposed iterative approach is that it allows one to evaluate
some elements of the disposal concept before money is spent on any construction work.
This applies in particular to the features, such as the materials to use in the construction of
the repository, the engineered barriers and the waste packages, whose physical and chemi-
cal properties are well understood, or can be determined in the laboratory. The influence
that different geometries of the repository, engineered barriers and waste packages may
have on the regulatory compliance criteria can also be evaluated at this stage and a few
selected for further investigation. The possibility that the system may fail and the conditions
under which a failure will occur can also be investigated and used to estimate the corre-
sponding nuclear liabilities, which can then be used to refine the decision analysis. This
approach will ensure that only those components and their properties that are really necessary
to ensure regulatory compliance will be included in the concept design. This can cause a con-
siderable potential reduction in the nuclear liabilities, the construction and assessment costs.
A post-closure assessment ends technically the moment a disposal system is licensed and
in operation. However, it would be foolish to stop all investigations of the system at that
stage and not proceed with a pre-closure assessment, which the discussion in Chapter 1
indicates should form an integral part of the post-closure assessment. One reason why this
may be necessary is that it is very difficult to include the consequences of a vehicle or other
accident on the site that may affect the post-closure assessment adversely. This will clearly
lengthen the institutional control period of the site and add to its nuclear liabilities, but one can
once again try to reduce this with a detailed decision analysis of the accident's consequences.
Any waste disposal system has a fini te life span, determined by the capacity of the repository,
at which time the system has to be closed and decommissioned. As used here the term
decommissioned refers to the actions that need to be taken, after its closure, with adequate
regard for the health and safety of workers and members of the public and the protection of
the environment (IAEA, 1993b). In the case of a disposal system this involves the orderly
Conclusions And Recommendations 144
decontamination, dismantling and removal of all the surface and subsurface support facilities,
erected during its operational life, the back-filling and sealing of all tunnels, shafts, exploratory
and monitoring boreholes and possibly the restoration of the site as close as possible to its
original environmental state. Although the repository will have to be closed at a specific
moment in time, it may sometimes be advantageous to delay the decommissioning of site,
especially when the results of a post-closure assessment is affected negatively by an accident
on the site. However, one should be able to control the intensity and duration of the
observations and the associated nuclear liabilities by a detailed analysis of the large volume
of information on the site's behaviour that should be available at that time.
10.3 RECOMMENDATIONS
A post-closure safety assessment of radioactive waste disposal systems is an extensive exer-
cise that requires input from various scientific, social and economic disciplines. No attempt
was therefore made to describe the actual implementation in detail. However, it is believed
that the thesis lie the foundation for a post-closure assessment procedure that will be within
the financial abilities of South Africa, and able to clarify the post-closure behaviour of
Vaalputs and to select a disposal site for the accumulated spent fuel. The same methodology
can also be used to decide whether Thabana should be closed permanently and cleaned up,
or developed into a permanent disposal site for radioactive waste, that will not impose un-
I.. due burdens on future generations.
As mentioned in Chapter 1 a preliminary version of the methodology has already been used
to assess the suitability of the BOSS concept for the disposal of spent nuclear sources.
Although the methodology has been refined since that assessment, it is felt that the proce-
dure used to generate scenarios for the assessment of uncertainties in the future evolution of
the site can be improved. No attempt was made to incorporate the decision analysis frame-
work, introduced for the first time in this thesis, in a consistent and logical manner into the
safety assessment methodology.
The safety assessment methodology need not be restricted to the waste disposal system as
such, but can also be used by regulatory authorities to justify their decisions to the owner or
operator of the disposal site and the public (Kozak, 1994). This includes aspects, such as
granting a licence for a specific site, or an order to close the site, restrictions placed on land
usage after the end of the institutional control period, and the durability requirements for
intrusion resistant covers (Siraky, 1996).
APPENDIX A
THE INTERNATIONAL ISAM LIST OF FEATURES, EVENTS AND
PROCESSES (FEPS) FORA RADIOACTIVE DISPOSAL SYSTEM
A.l DEFINITION OF THE DISPOSAL SYSTEM DOMAIN
Layers 1,2 and 30fthis international list are all defined relative to the following definition of the Disposal
System Domain.
The disposal system domain is defined as the spatial and temporal domain that consists of:
(a) the wastes,
(b) the engineered and natural barriers expected to contain the waste,
(c) the potentially contaminated geology and surface environment,
(d) the geology, surface environment and human behaviour necessary to provide an estimate of the
movement and the exposure of man to the radionuclides, following the closure of the repository.
A.2 LAYERS AND CATEGORIES OF THE LIST
Layer 0 Assessment Context
Assessment context factors are factors that the analyst will consider in determining the scope of the analysis;
these may include factors related to regulatory requirements, definition of desired calculation end-points and
requirements in a particular phase of assessment. Decisions at this point will affect the phenomenological
scope of a particular phase of assessment; i.e. what "physical FEPs" will be included. For example, some
classes of future human actions or extreme future events unrelated to the repository may be excluded.
Layer 1 External Factors
External Factors are FEPs with causes or origins outside the disposal system domain, i.e. natural or human
factors of a more global nature and their immediate effects. Included in this layer are decisions related to
repository design, operation and closure since these are outside the temporal bound of the disposal system
domain.
In general, external factors are not influenced, or only weakly influenced, by processes within the disposal
system domain. In developing models of the disposal system domain, external factors are often represented as
boundary conditions or initiating events for processes within the disposal system domain. The following
categories are used for this purpose.
1.1 Repository issues - Decisions on the design, waste allocation, and events related to the site
investigation, operations and closure.
1.2 Geological processes and effects - Processes arising from the wider geological setting and long-
term processes.
1.3 Climatic processes and effects - Processes related to global climate change and associated regional
effects.
1.4 Future human actions - Human actions and regional practices in the post-closure period that can
potentially affect the performance of the engineered and geological barriers, e.g. intrusive actions,
but not the passive behaviour and habits of the local population in 2.4 below.
1.5 Other - all other FEPs not included in 1.1to 1.4,e.g. meteorite impacts.
-145-
Appendix A 146
Note that there are only a few significant direct interactions between FEPs in the different categories of
external factors, in general.
Layer 2 Environmental Factors Affecting the Disposal System Domain
The Disposal system domain environmental factors are FEPs processes that occur within the spatial and
temporal domain, relevant to the estimation of the release and migration of radionuclides, and the effect that
has on the evolution of the physical, chemical, biological and human conditions, controlling the exposure of
man to the nuclides. (See also Layer 3). The following categories belong to this layer.
2.1 Waste and engineered barriers - The type~ structure and nature of these components.
2.2 Geological environment - The hydrogeological, geomechanical and geochemical FEPs related to
the pre-emplacement and modified state of the environment (and other long-term changes) caused by
the presence of the repository.
2.3 Surface environment - The FEPs related to surface phenomenon, including near-surface aquifers
and unconsolidated sediments, but excluding human activities and behaviour in 1.4 and 2.4.
2.4 Human behaviour - The habits and characteristics of individuals and population (in other words,
the critical group for which exposures have to be calculated) that will impact on the performance of
the engineered or geological barriers, but excluding the intrusive or other activities in 1.4.
The interactions between the FEPs in this layer of categories are clearly very important for the assessment.
Layer 3. Radionuclide and Contaminant Factors
The radionuclide and contaminant factors include those FEPs that will directly affect the disposal system
environment and the dose members of the critical group may receive from the given concentrations of
radionuclides in the environment. There are essentially three categories in this layer.
3.1 Contaminant characteristics - The characteristics of radiotoxic and chemotoxic nuclide species
that might be considered in a post-closure safety assessment.
3.2 Release and migration factors - The FEPs that directly affect the release and migration of
radionuclides in the disposal system domain.
3.3 Exposure factors - Processes and conditions that directly affect the dose members of the critical
group may receive from the given concentrations of radionuclides in the environmental media.
The boundaries between the different layers and categories are subjective and will depend on the individual
analyst's concept of the disposal system. Nevertheless, this should not prevent a self-consistent assignment of
FEPs within the International List itself or when mapping project FEPs to the International List.
A.3 THE DRAFT ISAM INTERNATIONAL FEP LIST (VERSION 1.0) IN CLASSIFICATION
SCHEME ORDER.
Layer 0 Assessment Context
0.01 Assessment endpoints
0.02 Time scales of concern
0.03 Spatial domain of concern
0.04 Repository assumptions
0.05 Future human action assumptions
0.06 Future human behaviour (target group) assumptions
0.07 Dose response assumptions
0.08 Assessment purpose
0.09 Regulatory requirements and exclusions
0.10 Model and data issues
Layer 1 External Factors
1.1 REPOSITORY ISSUES
1.1.01 Site investigation
Appendix A 147
1.1.02 Excavation and construction
1.1.03 Emplacement of wastes and back-filling
1.1.04 Closure e.g. capping
1.1.05 Records and markers, repository
1.1.06 Waste allocation
1.1.07 Repository design
1.1.08 Quality control
1.1.09 Schedule and planning
1.1.10 Administrative control, repository site
1.1.11 Monitoring of repository
1.1.12 Accidents and unplanned events
1.1.13 Retrievability
1.2 GEOLOGICAL PROCESSES AND EFFECTS
1.2.01 Tectonic movements and orogeny
1.2.02 Deformation, elastic, plastic or brittle
1.2.03 Seismicity
1.2.04 Volcanic and magmatic activity
1.2.05 Metamorphism
1.2.06 Hydrothermal activity
1.2.07 Erosion and sedimentation
1.2.08 Diagenesis
1.2.09 Salt diapirism and dissolution
1.2.10 Hydrological and hydrogeological response to geological changes
1.3 CLIMATIC PROCESSES AND EFFECTS
1.3.01 Climate change, global
1.3.02 Climate change, regional and local .
1.3.03 Sea level change
1.3.04 Periglacial effects
1.3.05 Glacial and ice sheet effects, local
1.3.06 Warm climate effects (tropical and desert)
1.3.07 Hydrological and hydrogeological response to climate changes
1.3.08 Ecological response to climate changes
1.3.09 Human response to climate changes
1.3.10 Other geomorphological changes
1.4 RITURE HUMAN ACTIONS
1.4.01 Human influences on climate
1.4.02 Motivation and knowledge issues (inadvertent and deliberate human actions)
1.4.03 Un-intrusive site investigation
1.4.04 Drilling activities (human intrusion)
1.4.05 Mining and other underground activities (human intrusion)
1.4.06 Surface environment, human activities
1.4.06.01 Surface Excavations
1.4.06.02 Pollution
1.4.06.03 Site Development
1.4.06.04 Archaeology
1.4.07 Water management (wells, reservoirs, dams)
1.4.08 Social and institutional developments
1.4.09 Technological developments
1.4.10 Remedial actions
1.4.11 Explosions and crashes
1.5 OTHER
1.5.01 Meteorite impact
1.5.02 Miscellaneous and FEPs of uncertain relevance
Appendix A 148
Layer 2 Environmental Factors Affecting the Disposal System Domain
2.1 WASTES AND ENGINEERED FEATURES
2.1.01 Inventory, radionuclide and other material
2.1.02 Waste form materials and characteristics
2.1.03 Container materials and characteristics
2.1.04 Buffer and backfill materials and characteristics
2.1.05 Engineered barriers system e.g. caps
2.1.06 Other engineered features materials and characteristics
2.1.07 Mechanical processes and conditions (in wastes and EBS)
2.1.08 Hydraulic and hydrogeological processes and conditions (in wastes and EBS)
2.1.09 Chemical and geochemical processes and conditions (in wastes and EBS)
2.1.10 Biological and biochemical processes and conditions (in wastes and EBS)
2.1.11 Thermal processes and conditions (in wastes and EBS)
2.1.12 Gas sources and effects (in wastes and EBS)
2.1.13 Radiation effects (in wastes and EBS)
2.1.14 Nuclear criticality
2.1.15 Extraneous materials
2.2 GEOLOGICAL ENVIRONMENT
2.2.01 Disturbed zone, host lithology
2.2.02 Host lithology
2.2.03 Lithological units, other
2.2.04 Discontinuities, large scale (in geosphere)
2.2.05 Contaminant transport path characteristics (in geosphere)
2.2.06 Mechanical processes and conditions (in geosphere)
2.2.07 Hydraulic and hydrogeological processes and conditions (in geosphere)
2.2.08 Chemical and geochemical processes and conditions (in geosphere)
2.2.09 Biological and biochemical processes and conditions (in geosphere)
2.2.10 Thermal processes and conditions (in geosphere)
2.2.11 Gas sources and effects (in geosphere)
2.2.12 Undetected features (in geosphere)
2.2.13 Geological resources
2.3 SURFACE ENVIRONMENT
2.3.01 Topography and morphology
2.3.02 Soil and sediment
2.3.03 Aquifers and water-bearing features, near surface
2.3.04 Lakes, rivers, streams and springs
2.3.05 Coastal features
2.3.06 Marine features
2.3.07 Atmosphere
2.3.08 Vegetation
2.3.09 Animal populations
2.3.10 Meteorology
2.3.11 Hydrological regime and water balance (near-surface)
2.3.12 Erosion and deposition
2.3.13 Ecological, biological and microbial systems
2.3.14 Animal and plant intrusion leading to vault and trench disruption
2.4 HUMAN BEHAVIOUR
2.4.01 Human characteristics (physiology, metabolism)
2.4.02 Adults, children, infants and other variations
2.4.03 Diet and fluid intake
2.4.04 Habits (non-diet-related behaviour)
2.4.05 Corrununity characteristics
2.4.06 Food and water processing and preparation
Appendix A 149
2.4.07 Dwellings
2.4.08 Wild and natural land and water use
2.4.09 Rural and agriculturailand and water use (including. fisheries)
2.4.10 Urban and industrial land and water use
2.4.11 Leisure and other uses of environment
Layer 3. Radionuclide and Contaminant Factors
3.1 CONTAMINANT CHARACTERISTICS
3.1.01 Radioactive decay and in-growth
3.1.02 Chemical and organic toxin stability
3.1.03 Inorganic solids and solutes
3.1.04 Volatiles and potential for volatility
3.1.05 Organics and potential for organic forms
3.1.06 Noble gases
3.2 CONTAMINANT RELEASE and MIGRATION FACTORS
3.2.01 Dissolution, precipitation and crystallisation, contaminant
3.2.02 Speciation and solubility, contaminant
3.2.03 Sorption and desorption processes, contaminant
3.2.04 Colloids, contaminant interactions and transport with
3.2.05 Chemical and complexing agents, effects on contaminant speciation and transport
3.2.06 Microbial, biological and plant-mediated processes, contaminant
3.2.07 Water-mediated transport of contaminants
3.2.08 Solid-mediated transport of contaminants
3.2.09 Gas-mediated transport of contaminants
3.2.10 Atmospheric transport of contaminants
3.2.11 Animal, plant and microbe mediated transport of contaminants
3.2.12 Human-action-mediated transport of contaminants
3.2.l3 Food chains, uptake of contaminants in
3.3 EXPOSURE FACTORS
3.3.01 Drinking water, foodstuffs and drugs, contaminant concentrations in
3.3.02 Environmental media, contaminant concentrations in
3.3.03 Non-food products, contaminant concentrations in
3.3.04 Exposure modes
3.3.05 Dosimetry
3.3.06 Radiological toxicity and other effects
3.3.07 Non-radiological toxicity and other effects
3.3.08 Radon and radon daughter exposure
REFERENCES
AEC (I997a) Solid Radioactive Waste Classification Scheme for theAEC. NWM-PRG-OOOI (Revision 1).
Nuclear Waste Management. Atomic Energy Corporation of SA Ltd., P.O. Box 582, Pretoria 0001.
AEC (I997b) Waste Acceptance Criteria for Vaalputs. NL28-NWM-WA-OI (Revision 4). Nuclear Waste
Management. Atomic Energy Corporation of SA Ltd., P.O. Box 582, Pretoria 0001.
AECB (1987) Regulatory Policy Statement. Regulatory Objectives, Requirements and Guidelines for the
Disposal of Radioactive Wastes - Long Term Aspects. Document R-I04. Atomic Energy Control Board
Regulatory.
AECL (1994a) Environmental Impact Statement on the Concept for Disposal of Canada's Nuclear Fuel
Waste. AECL-I07I1 Report, COG-93-I. Atomic Energy of Canada Ltd.
AECL (I994b) Summary of the Environmental Impact Statement on the Concept for Disposal of Canada's
Nuclear Fuel Waste. AECL-I0721 Report, COG-93-Il. Atomic Energy of Canada Ltd.
AECL (1997). Analysis of Safety Issuesfor the Preliminary Safety Analysis Report on the Intrusion Resistant
Underground Structure. AECL-MISC-386 Report. Atomic Energy of Canada Ltd.
Agg, P.J., Moreton, A.D., Rees, J.H., Rodwell, W.R., and Sumner, P.J. (1993) NSARP Reference Document,
Gas Generation and Migration, January 1992. Nirex Report 313, NSS/GI20. AEA Technology, Harwell,
Al-Mughrabi, M. (1998). Technical Manual for Conditioning of Spent Radium Sources. Working document
produced by IAEA.
Anderson, N .l.B. (1996) A Structural Analysis of the Vaalputs Waste Disposal Site and Environs, Bushmanland.
South Africa. Unpublished M.Sc. Thesis. Faculty of Science, University of Pretoria.
Andrade, C., Sagrera, J.L. and Alonso, C. (1997) Behaviour of Engineered Barriers. Lectured delivered at
the IAEA Interregional Training Course, Madrid, Spain.
Apted, MJ. and Engel, D.W. (1991) Validation of source-term models using natural analogues. In: Proceed-
ings of the Second Annual International High Level Radioactive Waste Management Conference. Las
Vegas, Nevada, Volume II: 1061-1065.
Bear, J. (1979) Hydraulics of Groundwater. Water Resources and Environmental Engineering. McGraw-
Hill, Book Co., New York.
Bear, 1. (1972) Dynamics of Fluids in Porous Media. American Elsevier Environmental Science Series.
American Elsevier Publishing Company, Inc., New York.
BIOMOVS II (1996) Development of a Reference Biospheres Methodology for Radioactive Waste Disposal.
BIOMOVS II Technical Report No. 6, published on behalf of the BIOMOVS II Steering Committee by
the Swedish Radiation Protection Institute, Sweden.
Bónáno, E.I., Hora, S.C., Keeney, R.L. and von Winterfeldt, D. (1990) Elicitation and Use of Expert Judge-
ment in Performance Assessment for High-level Radioactive Waste Repositories. NUREG/CR-541I,
SAND89-1821, Sandia National Laboratories, Albuquerque, NM 87185.
Botha, J. F. (1998) An Analytical Modelfor the Transport of Radioactive Material through the Earth's Sub-
surface. Report to the Environmental Consultant Anglogold-Vaal River Operations. Institute for
Groundwater Studies, University of the Orange Free State, P.O. Box 339, Bloemfontein 9300.
Botha, J. F. (1996) Principles of Ground water Motion. Unpublished Lecture Notes. Institute for Groundwater
Studies, University of the Orange Free State, P.O. Box 339, Bloemfontein 9300.
Botha, J. F. (1994) Models and The Theory of Groundwater Motion. Unpublished Report. Institute for
Groundwater Studies, University of the Orange Free State, P.O. Box 339, Bloemfontein.
Botha, J. F., Verwey, J. P., Van der Voort, I., Vivier, J. J. P., Colliston, W. P. and Loock, J. C. (1998) Karoo
Aquifers. Their Geology, Geometry and Physical Behaviour. WRC Report No 487/1/98. Water Research
Commission, P.O. Box 824, Pretoria 0001
Brewitz, W. (1986) Feasibility of disused underground mines for low level and intermediate level waste
disposal. Treatment and disposal of radioactive waste, and its disposal in arid environments. In: Proceed-
ings of the Radwaste '86 Conference, Cape Town, Atomic Energy Corporation of SA Ltd., Pretoria.
Brynard, HJ. (1989) A Pertrological and Geochemical Study Regarding the Suitability of the Vaalputs Ra-
dioactive Waste Disposal Facility. Unpublished Ph.D. Thesis. University of Pretoria.
-150-
References 151
Bugai, D., Smith, L. and Beckie, R. (1996) Risk-cost analysis of strontium-90 migration to water wells at the
Chernobyl Nuclear Power Plant. Environmental & Engineering Geosciences, 2, (2), 151-164.
CarIsson, J., Papp, T. and Forsstrëm, M. (1990) Safety Assessment of SFR - A Swedish repository for low
and intermediate level radioactive waste. In: Proceedings of the Symposium on the Safety Assessment of
Radioactive Waste Repositories. OEDC, Paris.
Chapman, N.A. and McKinley, l.G. (1988) The Geological Disposal of Nuclear Waste. John Wiley & Sons
Ltd.
Chapman, N.A., Andersson, J., Robinson, P., Skagius, K., Wene, C-O., Wiborgh, M. and Wingefors, S.
(1995). Systems Analysis, Scenario Construction and Consequence Analysis Definitionfor SITE-94. Swed-
ish Nuclear Power Inspectorate Report No. 95:26, Stockholm, Sweden.
Cho, W., Chang, S. and Park, H. (1992) Uncertainty analysis of safety assessment for high-level radioactive
waste repository. Waste Management. 12,45-54.
Corner, B. and Scott, J.R. (1980) Report on the Screening Phase. Site Selection Program for the Disposaf/
Storage of Radioactive Waste in South Africa. PER-lOO. (GEA-287). Atomic Energy Corporation of SA
Ltd., Pretoria.
Courtnage, P.M, Andreoli, M.A.G and Heard, R.G, (1996) Post-Transvaal Deformation Between the Johan-
nesburg Dome and the Bushveld Complex, with Particular Emphasis on the Geology of Pelindaba. GEA-
1186. Atomic Energy Corporation. of S.A. Ltd.
Cranwell, R.M., Guzowski, R.W., Campbell, J.E., and Ortiz, N.R. (1990). Risk Methodology for Geological
Disposal of Radioactive Waste: Scenario Selection Procedures. NUREG/CR-I667. SAND80-1429. Sandia
National Laboratories, Albuquerque, NM 87185.
Crouch, E.A.C. and Wilson, R. (1982) Risk/Benefit Analysis. Ballinger, Baston, MA.
Darcy, H. P. G. (1856) Les Fontaines Publiques de la Ville de Dijon. Dalmont, Paris.
De Marsily, G., Combes, P. and Goblet, P. (1992) Comment on 'Ground-water models cannot be validated',
by L.F. Konikow & J.D. Bredehoeft. Advances in Water Resources. 15,367-369.
Dolinar, G.M., J.H. Rowat, M.E. Stephens (1996) Preliminary Safety Analysis Report (PSAR) for the Intrusion
Resistant Underground Structure. AECL-MISC-295 Report (Revision 4). Atomic Energy of Canada Ltd ..
Easterling, D. And Kunreuther, H. (1995) The Dilemma of Siting a High-Level Nuclear Waste Repository.
Kluwer Academic Publishers, Boston.
Ebel, K. and Richter, D. (1986) Experience from developing and licensing an underground repository for
low and intermediate level wastes. In: Proceedings of the Symposium on the Siting, Design and Con-
struction of Underground Repositories for Radioactive Wastes. IAEA, Vienna.
Ellis, J. F. (1999) The Assessment of Potential Radiation Hazards from Gold Mines in the Free State Gold-
fields to Members of the Public. Unpublished M.Sc. Thesis. Department of Medical Physics, University
of the Orange Free State, P.O. Box 339, Bloemfontein.
Environment Agency, Scottish Environment Protection Agency and Department of the Environment for
Northern Ireland (1997) Radioactive Substance Act 1993 - Disposal Facilities on Land for Low and
Intermediate Level Radioactive Waste: Guidance on Requirements for Authorisation. HMSO, London.
Eng, T., Hudson, J. Stephansson, 0., Skagius, K. and Wiborgh, M. (1994) Scenario development methodolo-
gies. SKB Technical Report 94-28, Swedish Nuclear Fuel and Waste Management Company, Sweden.
Forsstrom. H. and Westerlind M. (1997) Non-TechnicaIAspects. Lecture delivered at the Workshop on Safety
Assessment Methodologies for Near Surface Disposal Facilities. Stockholm, Sweden.
Freeze, R.A. and Cherry, J.A. (1979) Groundwater. Prentice-Hall, Inc. Englewood Cliffs, New Jersey 07632.
Freeze, R.A., Smith, L.,Massman, J.w., Sperling, T. and James, B. (1990) Hydrogeological Decision Analy-
sis: 1. A Framework. Ground Water, 28, (5), 738-766.
Freeze, R.A., James, B., Massman, J .w., Sperling, T. and Smith, L. (1992) Hydrogeological Decision Analy-
sis: 4. The Concept of Data Worth and its Use in the Development of Site Investigation Strategies. Ground
Water. 30, (4), 574-588.
Gitrnan, LJ. (1988) Principles of Managerial Finance. Fifth edition. Harper & Row Publishers, New York.
Groebner, D.F. and Shannon, P.w. (1989) A decision-making approach. Merrill Publishing Company (3rd
edition), Columbus, Ohio.
Guzowski, R.V. (1990) Evaluation of applicability of probability techniques to determining the probability
of occurrence of potentially disruptive intrusive events at the WIPP. SAND90-7100, Sandia National
Laboratories, Albuquerque, NM 87185.
Guzowski, R. V., and G. Newman, 1993. Preliminary Identification of Potentially Disruptive Scenarios at
the Greater Confinement Disposal Facility, Area 5 of the Nevada Test Site. SAND93-7100, Sandia Na-
tional Laboratories, Albuquerque, NM 87185.
References 152
Hambleton-Jones, B.B., Snyders, L.S. and Heard, R.G. (1998) The Management of Radioactive Waste in
South Africa. Poster presented at The National Nuclear Technology Conference. The University of North
West, Mmabatho, South Africa.
Holtz, G. (1986) Technical design of the Konrad repository. In: Proceedings of the Symposium on the Siting,
Design and Construction of Underground Repositories for Radioactive Wastes. IAEA, Vienna.
Hollister, C. D. and Nadis, S. (1998) Burial of radioactive waste under the seabed. Scientific American. 278
(1),40-45.
Hsieh, P.A., Neuman, S. P. and Simpson, E. S. (1983) Pressure Testing of Fractured Rocks -A Methodology
Employing Three-dimensional Cross-hole Tests. Topical Report. Report No. NUREG/CR-3213RW. De-
partment of Hydrology and Water Resources, Univ. of Arizona, Tucson, AZ 85721.
IAEA (1981) Safety Assessment for the Underground Disposal of Radioactive Waste. International Atomic
Energy Agency Safety Series. No. 56. International Atomic Energy Agency, Vienna.
IAEA (1983) Criteria for Underground Disposal of Solid Radioactive Waste. International Atomic Energy
Agency Safety Series. No. 60. International Atomic Energy Agency, Vienna.
IAEA (1989) Safety Principles and Technical Criteriafor the Underground Disposal of High Level Radioac-
tive Wastes. International Atomic Energy Agency Safety Series. No. 99. International Atomic Energy
Agency, Vienna.
IAEA (1990) Sealing of Underground Repositories for Radioactive Wastes. International Atomic Energy
Agency Technical Reports Series. No. 319. International Atomic Energy Agency, Vienna.
IAEA (1991). Nature and Magnitude of the Problem of Spent Radiation Sources. IAEA-TECDOC-620.
International Atomic Energy Agency, Vienna.
IAEA (1992) Radioactive Waste Management. An IAEA Source Book. International Atomic Energy Agency,
Vienna.
!AEA (l993a) Radioactive Waste Management Glossary. 3rd ed, International Atomic Energy Agency, Vienna.
IAEA (1993b) Report on Radioactive Waste Disposal. IAEA Technical Report Series. No. 349. International
Atomic Energy Agency, Vienna.
IAEA (1994a) Classification of Radioactive Waste. International Atomic Energy Agency Safety Series. No.
III-G-I.l. International Atomic Energy Agency, Vienna.
IAEA (1994b) Siting of Geological Disposal Facilities. International Atomic Energy Agency Safety Series.
No. III-G-4.I. International Atomic Energy Agency, Vienna.
IAEA (I994c) Siting of Near Surface Disposal Facilities. International Atomic Energy Agency Safety Se-
ries. No. III-G-3.1. International Atomic Energy Agency, Vienna.
IAEA (1995a) The Principles of Radioactive Waste Management. International Atomic Energy Agency Safety
Series. No. lII-F. International Atomic Energy Agency, Vienna.
IAEA (l995b) International Safety Standards for Protection against Radiation andfor the Safety of Radia-
tion Sources. Basic Safety Standards 115. International Atomic Energy Agency, Vienna.
IAEA (l995c) Establishing a National System for Radioactive Waste Management. International Atomic
Energy Agency Safety Series. No. lII-S-I. International Atomic Energy Agency, Vienna.
IAEA (1996). Conditioning and Interim Storage of Spent Radium Sources. IAEA-TECDOC-886. Interna-
tional Atomic Energy Agency, Vienna.
IAEA (1997a). The International Programme for Improving Long Term Safety Assessment Methodologies
for Near Surface Radioactive Waste Disposal Facilities: Objectives, Content and Work Programme. ISAM
Document ISAM/GIOI97. International Atomic Energy Agency, Vienna.
IAEA (I997b) Safety Indicators in Different Time Frames for the Safety Assessment of Underground Radio-
active Waste Repositories. IAEA- TECDOC-767. International Atomic Energy Agency, Vienna.
IAEA (1998a) International Review of the Radiological Situation at the Vaalputs National Radioactive Waste
Management and Disposal Facility, South Africa. Unpublished IAEA Report. International Atomic En-
ergy Agency, Vienna.
IAEA (1998b) Development of an Information System for Features, Events and Processes (FEPs) and Ge-
neric Scenarios for Safety Assessment of Near Surface Radioactive Waste Disposal Facilities, ISAM
Document ISAM/SWGI0298. Version 0.2. International Atomic Energy Agency, Vienna.
IAEA (1998c) Factorsfor Formulating Strategies for Environmental Restoration. IAEA- TECDOC-I032.
International Atomic Energy Agency, Vienna.
IAEA (1998d) Characterization of Radioactively Contaminated Site for Remediation Purposes. IAEA-
TECDOC-I032. International Atomic Energy Agency, Vienna.
IAEA (1999) Derivation of Quantitative Acceptance Criteria for Disposal of Radioactive Waste to Near
Surface Facilities: Development and Implementation of an Approach. Draft safety report, Working docu-
References 153
ment. Version 3.0. International Atomic Energy Agency, Vienna.
ICRP (1977) Recommendations of the ICRP. ICRP Publication 26. Annals of the ICRP 1(3), Pergamon
Press, Oxford.
ICRP (1990) 1990 Recommendauons of the International Commission on Radiological Protection. ICRP
Publication 60. Pergamon Press, Oxford.
ICRP (1997) Radiological Protection Policy for the Disposal of Radioactive Waste. ICRP Publication 77.
Pergamon Press, Oxford.
Iman, RL. and Shortencarier, MJ. (1984) A Fortran 77 Program and User's Guide for the Generation of
Latin Hypercube and Random Samples for use with Computer Models. NUREG/CR-3624, SAND83-
2365, Sandia National Laboratories, Albuquerque, NM 87185.
Issler, H. (1990). Waste from medicine, industry and research must also be disposed of. In: NAGRA Bulletin.
1/1990.
James, B.R. and Gorelick, S.M. (1994) When enough is enough: The worth of monitoring data in aquifer
remediation design. Water Resources Res. 30, (12), 3499-3514.
James, B.R. and Freeze, R.A. (1993) The worth of data in predicting aquitard continuity in hydrogeological
design. Water Resources Res. 29, (7), 2049-2065.
James, B.R., Gwo, J. and Toran, L. (1996a) Risk-Cost Decision Framework for Aquifer Remediation De-
sign. Journal of Water Resources Planning and Management. 122, (6), 414-419.
James, B.R., Huff, D.D., Trabalka, J.R., Ketelle, R.H. and Rightmire, C.T. (1996b) Allocation of Environ-
mental Remediation Funds Using Economic Risk-Cost-Benefit Analysis: A Case Study. Ground Water
Management and Remediation (GWMR). 95-105.
Janse van Rensburg, HJ. (1992) Geohidrologiese Besluitnemingsanalise- 'n Koste-, Voordeel- en
Risikobenadering. (In Afrikaans). Unpublished. MBA Thesis. UNISA, Pretoria, South Africa.
Jardine, K., Smith, L. and Clerno, T. (1996) Monitoring Network in Fractured Rocks: A Decision Analysis
Approach. Ground Water, 34, (3),504-518.
Kennedy, W.E. (1997) General perspective on performance assessment. In: Proceedings of the 18th Annual
USDOE Low-Level Radioactive Waste Management Conference. Salt Lake City, Utah, USA.
Kickmaier, W. and McKinley, I. (1996) Investigations in Phase I to III (1983 - 1993) atthe Grimsel Test Site
-A Retrospective. In: NAGRA Bulletin No. 27.
Kirkwood, C.w. (1997) Strategic Decision Making. Multiobjective decision analysis with spreadsheets.
Duxbury Press, Belmont, California.
Konikow, L. F. and Bredehoeft, J. D. (1992) Ground-water models cannot be validated. Advances in Water
Resources. 15, 75-83.
Kozak, M.W. (1994) Decision analysis for low-level radioactive waste disposal safety assessments. Radio-
active Waste Management and Environmental Restoration, 18, 209-223.
Kozak, M.W. (1997) Sensitivity, uncertainty, and importance analyses. In: Proceedings of the 18th Annual
USDOE Low-Level Radioactive Waste Management Conference. Salt Lake City, Utah, USA.
Kozak, M.W. (1998) Personal Communication.
Kozak, M.W. and Zhou, W. (1998) The use of interaction matrices to improve assessment transparency. In:
Alternative Approaches to Assessing the Performance and Suitability of Yucca Mountain for Spent Fuel
Disposal. Electric Power Research Institute Report TR-I08732, Palo Alto.
Kozak, M.W., Stenhouse, MJ., Van Blerk, JJ. and Heard, RG. (1999) Borehole Disposal of Spent Sources,
Volwne II-Initial Safety Assessment and Evaluation of the Disposal Concept. GEA-1353. Atomic En-
ergy Corporation of SA Ltd., P.O. Box 582, Pretoria 0001.
Kozak, M.w., Olague, N.E., Rao, R.R. and McCord, J.T. (1993) Evaluation of a Performance Assessment
Methodology for Low-Level Radioactive Waste Disposal Facilities. Evaluation of Modelling Approaches.
NUREG/CR-5927, SAND91-2802, Vol. 1. Sandia National Laboratories, Albuquerque, NM 87185.
Levin, M. (1988) A Geohydrological Appraisal of the Vaalputs Radioactive Waste Disposal Facility in
Namaqualand, South Africa. Unpublished Ph.D. Thesis, Faculty of Science, Department of Geohydrology,
University of the Orange Free State, P.O. Box 339, Bloemfontein.
Massman, J. and Freeze, RA. (1987a) Groundwater contamination from waste management sites: The inter-
action between risk-based engineering design and regulatory policy. 1.Methodology. Water Resources
Res. 23. 351-367.
Massman, J. and Freeze, R.A. (1987 b) Groundwater contamination from waste management sites: The inter-
action between risk-based engineering design and regulatory policy. 2. Results. Water Resources Res. 23.
368-380.
Massman, J., Freeze, R.A., Smith, L., Sperling, T. and James, B. (1991) Hydrogeological decision analysis:
References 154
2. Application to groundwater contamination. Ground Water, 29, (4), 537-548.
Moore, P.E., Adrian, H.W.W., Bain, c.A.R., Faurie, J.N., Gerber, H.H., Levin, M., van As, D., Oelofse, P.J.,
Posnik, S.J., Hambleton-Jones, B.B., Venter, P.W.J., Van der Westhuizen, HJ. (1987) Vaalputs National
Radioactive Waste Disposal Safety Report. PIN-lOlO. Atomic Energy Corporation of SA Ltd., P.O. Box
582, Pretoria 0001.
Moore, P.E. and Hambleton-Jones, B.B. (1996) Planning and operation ofVaalputs, Tbe South African Na-
tional Radioactive Waste Management Facility. In: Proceedings of the Symposium on the Planning and
Operation of Low Level Waste DisposalFacilities. International Atomic Energy Agency, Vienna.
Niebuhr, H.W., R.F. Henwood, and J.R. Colley (1968) Report on the Permanent Storage Area for Active
Solid Waste. (Previously the Chemical Operations Division, Atomic Energy Board), Atomic Energy Cor-
poration of SA Ltd., P.O. Box 582, Pretoria 0001.
OECD/NEA (1977) Objectives, Concepts and Strategies for the Management of Radioactive Waste Arising
from Nuclear Power Programmes. Report by a Group of Experts of the OECD/NEA. Organisation for
Economic Co-operation and Development, Paris.
OECD/NEA (1993). The Cost of High-Level Waste Disposal in Geological Repositories. An Analysis of
Factors Affecting Cost Estimates. Organisation for Economic Co-operation and Development, Paris.
OECD/NEA (1996). Future Financial Liabilities of Nuclear Activities. Organisation for Economic Co-op-
eration and Development, Paris.
OECD/NEA (1998). Safety Assessment of Radioactive Waste Repositories: An International Database of
Features, Events and Processes. A report of the NEA Working Group on the development of a database of
features, events and processes relevant to the assessment of post-closure safety of radioactive waste re-
positories. Organisation for Economic Co-operation and Development, Paris.
Price, L.L., Conrad, S.H., Zirnmerman, D.A., Olague, N.E., Gaither, K.C., Cox, W.B., McCord, J.T. and
Harlan, C.P. (1996) Preliminary Performance Assessment of the Greater Confinement Disposal Facility
at the Nevada Test Site. Volume 1. Excecitive Summary. SAND91-0047. Sandia National Laboratories,
Albuquerque, NM 87185. .
Phillip, J. and Cliftin, J.R. (1989) Durability of concrete for underground containment ofLLW. In: Proceed-
ings of the Symposium on Waste Management '89. Waste Processing, Transportation, Storage and Dis-
posal. Volume II. Low Level Waste. Tucson, Arizona, USA. .
Redding, S. and Hutson, J.L. (1983) Part I: General Climate of Bushmanland. Part II: Estimates of Percola-
tion at the Buried Waste Facility, Bushmanland. PER-114 (GEA-636). Atomic Energy Corporation of SA
Ltd., P.O. Box 582, Pretoria 0001.
Quigley, R. M. and Fernandez, F. (1989) Clay/organic interactions and their effect on the hydraulic conduc-
tivity of barrier clays. In: Proceedings of the IAHR Symposium: Contaminant Transport in Groundwater.
H. E. Kobus and W. Kinzelbach (eds.) Stuttgart. A.A. Balkema, Rotterdam.
Skrable, K., French, C., Chabot, G. and Major, A. (1974) A general equation for the kinetics of linear first
order phenomena and suggested applications. Health Physics (27), 155-157.
Seitz, R.R. Garcia, R.S. Kostelnik, K.M. and Starmer, R.J. (1992). Performance Assessment Handbookfor
Low-Level Radioactive Waste Disposal Facilities. DOE-LLW-135. Idaho National Engineering Labora-
tory, Idaho Falls, Idaho 83415.
Shuman, R. and Rogers, V.C. (1989) The BARRIER code for predicting long-term concrete performance. In:
Proceedings of the Symposium on Waste Management '89. Waste Processing, Transportation, Storage
and Disposal. Volume II. Low Level Waste. Tucson, Arizona, USA.
Siraky, G. (1996) Safety Analysis as a Regulatory Tool. In: Proceedings of the Symposium on the Planning
and Operation of Low Level Waste Disposal Facilities. International Atomic Energy Agency, Vienna.
Skagius, K., Strëm A. and Wiborgh, M. (1995). The Use of Interaction Matricesfor Identification, Structur-
ing and Ranking of FEPs in a Repository System: Application on the Far-field of a Deep Geological
Repository for Spent Fuel. SKB Technical Report 95-22. Stockholm, Sweden.
Skagius, K., and Wiborgh, M. (1994) Testing of Influence Diagrams as a Tooifor Scenario Development by
Application on the SFL 3-5 Repository Concept. SKB Technical Report 94-47. Swedish Nuclear Fuel and
Waste Management Company.
Smith, G.M., B.M. Watkins, R.H. Little, H.M. Jones, and A.M. Mortimer (1996) Biosphere Modelling and
Dose Assessment for Yucca Mountain, EPRI TR-I07190. Electric Power Research Institute, Palo Alto.
Smith, P., Zuiderna, P. and McKinley, l.G. (1994) Assessing the performance of the NAGRA HLW disposal
concept. In: NAGRA Bulletin. 25/1994.
Sperling, T., Freeze, R.A., Massman, J., Smith, L., and James, B. (1992) Hydrogeological decision analysis:
3. Application to design of a ground-water control system at an open pit mine. Ground Water, 30, (3), 376-
References 155
389.
Sudicky, E. A. (1983) An Advection-Diffusion Theory of Contaminant Transportfor Stratified Porous Media.
Unpublished Ph.D. Thesis. Department of Earth Sciences, University of Waterloo, Ontario.
Sullivan, T.M. and Kozak, M.W. (1996) Contaminant Transport in Groundwater. Lecture delivered at the
IAEA Interregional Training Course on Near Surface Disposal of Low and Intermediate Level Waste,
Cumbria, UK.
Sullivan, T.M. (1996) DUST-MS - Model Equations for Waste Form Leaching and Transport with Ingrowth
Due to Radioactive Decay. Environmental & Waste Technology Center, Brookhaven National Labora-
tory.
Sullivan, T.M. (1997) Modelling of Source Term. Lecture delivered at the IAEA Interregional Training Course
on Safety Assessment Methodologies for Near-Surface Radioactive Waste Disposal Facilities, Madrid,
Spain.
Sumerling, TJ., D. Charles, D.P. Hodgkinson, and G.M. Smith (1989) The Development and Specification of
Scenarios for Post-Closure Radiological Safety Assessment of the Drigg LLW Disposal Site. Intera- Ex-
ploration Consultants Limited (now QuantiSci) report 11931-2 (Version 4).
Torres, C. and Simon, I. (1997) Biosphere Transport and Dose Modelling. Lecture delivered at the lAEA
Interregional Training Course on Safety Assessment Methodologies for Near-Surface Radioactive Waste
Disposal Facilities, Madrid, Spain.
USDOE (1988) Site Characterization Plan, Yucca Mountain Site. DOEfRW-198 Report. United States De-
partment of Energy, Washington, DC.
USDOE (1990) Final Supplement - Environmental Impact Statement: Waste Isolation Pilot Plant. DOE.ETS-
0026-FS Report. Office of Environmental Restoration and Waste Management, United States Department
of Energy, Washington, DC.
USDOE (1998). "The Nuclear Age Line". Center for Environmental Management United States Department
of Energy. http://www.em.doe.gov/timeline.
Van As, D. and Basson, J.K. (1969) Study of radioactivity in the environment of the national nuclear research
centre, Pelindaba. South African Journal of Science, 65(1), 8-15.
Van Blerk,JJ., Vivier, J.J.P., Pirow, P., Andreoli, M.A.G. and Heard, R.G. (1999) Borehole Disposal of Spent
Sources. Volume I - Development of the Concept. GEA-1353 Report. Nuclear Waste Systems, Atomic
Energy Corporation Ltd., P.O. Box 582, Pretoria 0001.
Van der Westhuizen, HJ. and Grobbelaar, N. (1967) Ioonabsorpsie-Eienskappe van die Grond by die
Bergingsterrein vir Vaste Radioaktiewe Afval. (In Afrikaans). Pe1150. Atomic Energy Corporation of SA
Ltd., P.O. Box 582, Pretoria 0001.
Van Sandwyk, L., Van Tonder, G. J., De Waal, D. J. and Botha, J. F. (1992)A Comparison of Spatia I Bayesian
Estimation and Classical Kriging Procedure. WRC Report No 271/3/92. Water Research Commission,
P.O. Box 824, Pretoria 0001.
Walton, J.C. Plan sky, L.E. and Smith, R.W. (1990) Models for Estimating of Service Life of Concrete Barri-
ers in Low-Level Radioactive Waste Disposal. NUREG/CR-5542, EGG-2597. Idaho National Engineer-
ing Laboratory, EG&G Idaho, Inc.
Walton, J.C. and Seitz R.R. (1991) Performance of Intact and Partially Degraded Concrete Barriers in
Limiting Fluid Flow. NUREG/CR-5614, EGG-2614. Idaho National Engineering Laboratory, EG&G Idaho,
Inc.
Watts, L. Clements, L., Egan, M., Chapman, N., Kane, P. and Thorne, M. (1999) Development of Scenarios
within a Systematic Assessment Framework for the Drigg Post-Closure Safety Case. Nuclear Energy
Agency Workshop on Scenario Development. Madrid, 10-12 May 1999.
Wolfram, S. (1996) The Mathematica Book. (3rd ed.) Wolfram Media/Cambridge University Press.
SUMMARY
Radioactive waste in South Africa is generated through the nuclear fuel cycle and the appli-
cation of radioactive materials in industry, science and medicine. The radioactive waste is
presently disposed at Vaalputs in Bushmanland and Thabana at Pelindaba in near-surface
disposal facilities. No strategy exists at present for the disposal of high level waste.
The objective of radioactive waste management and its underlying principles is to ensure
that human health and the environment are protected at all times, without imposing an
undue burden on future generations. This implies that, before any long-term management
strategy of radioactive waste disposal can be implemented, the impact of the disposed waste
must be determined as a function of time-a procedure referred to as post-closure safety
assessment.
In this thesis, a methodology to perform post-closure safety assessments of radioactive waste
disposal systems in South Africa and other parts of Africa is described. Not only will it
contribute significantly to reassess the suitability of current waste disposal practices, but
also lays the foundation for future disposal practices.
The proposed methodology-an integrated approach to radioactive waste management-is
aimed at:
(a) ensuring the safety of the present public and future generations,
(b) enhancing the public acceptance of the methodology,
(c) keeping the expenditure associated with the implementation of the methodology at
a minimum.
The methodology recognises the interdependence between operational phase activities and
the post-closure behaviour of the disposal system. It is an iterative process that considers
site-specific, prospective evaluations of the post-closure phase to ensure that the disposal
system will comply with internationally accepted criteria, within reasonable limits. Provi-
sion is therefore made to identify the data, design and other needs that will contribute to-
wards the achievement of this objective. The first step in this procedure is to identify those
internal and external features, events and processes that can be used to predict how radioac-
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Summary 157
tive material may escape from the disposal facility, along which paths will it migrate and
how it may impact humans. Various conceptual and mathematical models that can be used
to develop appropriate scenarios of these processes and to compare the results with site-
specific data are discussed in the thesis.
The cost to develop a waste disposal system, the disposal of the waste and the pre- and post-
closure assessments of the system, or so-called nuclear liabilities, can be astronomically
high. Combining the post-closure assessment ofthe system with the decision analysis frame-
work discussed in the thesis can reduce these costs considerably.
Post-closure assessments of radioactive waste disposal systems have in the past often been
interpreted as an exercise to predict the exact behaviour of the system far into the future.
However, as pointed out in the thesis this is not possible, even with the technology available
today. The more pragmatic approach, advanced in the thesis, is that modem technology is
able to demonstrate to reasonable members of the public that such a system will be safe.
Nevertheless it is recognized that the methodology cannot be implemented without the ac-
tive participation of the public. It is therefore envisaged that the proposed methodology will
be implemented with the close co-operation of the public, particularly those living near the
site where the disposal system will be implemented.
OPSOMMING
In Suid Afrika word radioaktiewe afval gegenereer deur die kernbrandstofsiklus en die gebruik
van radioaktiewe bronne in die industrie, die wetenskap en geneeskunde. Lae en middel
energie radioaktiewe afval word tans weggedoen in oppervlak wegdoeningsfasiliteite by
Vaalputs in die Boesmanland en geberg in vlak bergingsfasiliteite by Thabana by Pelindaba.
Geen strategie bestaan egter tans vir die wegdoening van hoë energie afval nie.
Die doel van radioaktiewe afvalbestuur en die onderliggende beginsels daarvan, is om te
verseker dat die mens se gesondheid asook die omgewing te alle tye beskerm word, sonder
om onnodige druk op toekomstige geslagte te plaas. Dit impliseer dat, alvorens enige
langtermyn bestuurstrategie vir radioaktiewe afval geïmplementeer kan word, moet die impak
van die gebergde afval as 'n funksie van tyd bepaal word-'n prosedure waarna verwys
word as 'n nasluiting veiligheidsvasstelling (post-closure safety assessment).
'n Metodiek om so 'n nasluiting veiligheidsvasstelling vir radioaktiewe afval
wegdoeningsisteme in Suid Afrika en ander dele van Afrika uit te voer, word in hierdie
proefskrif beskryf. Nie alleen bied dit 'n beduidende bydrae tot die herevaluering van
bestaande wegdoeningsisteme nie, maar verskaf dit ook 'n basis vir toekomstige
wegdoeningspraktyke wêreldwyd.
Die voorgestelde metodiek-'n geïntegreerde benadering tot radioaktiewe afvalbestuur-
het ten doelom:
(a) die veiligheid van huidige en toekomstige geslagte te verseker,
(b) publieke aanvaarding van die metodiek te bevorder,
(c) finansiële uitgawes geassosieer met die implementering van die metodiek te minimeer.
Die metodiek aanvaar die interafhanklikheid wat daar bestaan tussen aktiwiteite binne die
operasionale fase en die nasluitingsgedrag van die sisteem. Dit is 'n iteratiewe proses wat
bestaan uit terrein spesifieke, toekomstige evaluasies van die nasluiting fase, met die doel
om te verseker dat die wegdoeningsisteem voldoen aan internasionaal aanvaarbare kriteria,
gegee sekere redelike beperkings. Gevolglik word voorsiening gemaak vir die identifisering
van data, ontwerp en ander behoeftes wat 'n bydrae saliewer om die doelwit te bereik. Die
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Opsomming 159
eerste stap in hierdie prosedure is om alle interne en eksterne eienskappe, gebeurtenisse en
prosesse (features, events and processes) te bepaal wat gebruik kan word om te voorspel
hoe radioaktiewe materiaal vanuit die wegdoeningsfasiliteit kan ontsnap, langs watter paaie
dit sal beweeg en wat die gevolglike impak op die mens sal wees. Verskeie konsepsuele en
wiskundige modelle wat gebruik kan word om die prosesse voor te stel en die resultate met
terrein spesifieke data te vergelyk, word in die proefskrif bespreek.
Die koste om so 'n radioaktiewe afval wegdoeningsisteem te ontwikkel, die wegdoening
van die afval en die ondersoeke wat nodig is om vas te stelof so 'n sisteem veilig sal wees,
algemeen bekend as kernaanspreeklikheid, kan astronomies hoog wees. Soos aangetoon in
die proefskrifkan hierdie kostes egter aansienlik verminder word deur die nasluitingsevaluasie
van die sisteem met 'n besluitnemingsanalise raamwerk te kombineer.
Nasluitingsevaluasies van radioaktiewe afval wegdoeningsisteme is dikwels in die verlede
geïnterpreteer as ' n oefening wat daarop ingestel is om die werklike verre toekomstige
gedrag van die sisteem te voorspel. Soos in die proefskrif aangedui word, is dit egter nie
moontlik nie, self nie eers met die tegnologie wat vandag beskikbaar is nie. Wat die moderne
tegnologie wel in staat is om te doen, is om aan te toon dat so 'nsisteem binne aanvaarbare
perke veilig is. 'n Meer pragmatiese benadering word dus in die proefskrif voorgestel, nl.
om aan redelike persone aan te dui dat so ' n sisteem wel veilig sal wees. Die metodiek kan
dus nie geïmplementeer word sonder die aktiewe deelname van die publiek nie. Die sukses
van die voorgestelde metodiek sal dus hoofsaaklik bepaal word deur die samewerking van
die publiek, veral dié persone wat naby 'n terrein woon wat vir so 'n wegdoeningsisteem
geoormerk is.