Post-Closure Safety Assessment for
Borehole Disposal of Disused Sealed
Sources in Ghana
1
EXECUTIVE SUMMARY
Ghana has disused radioactive sources that need to be managed and disposed of
carefully and in a safe and secure manner. These sources contain different
radionuclides in highly variable quantities. Many sources are small in physical size,
however they can contain very high activities, with typical levels in the 7.4E+5 Bq to
6.85E+14 Bq range. Therefore, if they are not managed properly, these radioactive
sources can represent a significant hazard to human health and the environment.
Storage in a secure facility can be considered as an adequate final management
option for sources containing quantities of short-lived radionuclides, which decay to
harmless levels within a few years. However, for some other sources a suitable
disposal option is required.
. Deep geological disposal offers the highest level of isolation available within
disposal concepts currently actively considered. Such facilities are under
consideration for the disposal of spent nuclear fuel, high level waste and
intermediate level waste in a number of countries. However, they are expensive to
develop and only viable for countries with extensive nuclear power programmes. As
Ghana does not have an extensive nuclear power programme that would require the
construction of a deep geoogical disposal facility, the disposal of disused sources in
narrow diameter (a few tens of centimetres) borehole facilities would appear to
provide a safe and cost effective disposal option for its disused sources.
A variety of borehole designs have been used for the disposal of radioactive waste
with differing depths (a few metres to several hundred metres) and diameters (a few
tens of centimetres to several metres). The design evaluated in this report is based on
the narrow diameter (0.26 m) design developed under the International Atomic
Energy Agency’s (IAEA) AFRA project (see Figures A and B) since this design has
been developed specifically for the disposal of disused radioactive sources and uses
borehole drilling technology that is readily available in the country. The design can
accommodate disused sources of less than 110 mm in length and 15 mm in diameter
meaning that the design is applicable to a wide range of sources. The sources are to
be disposed at a depth of 56.5m from the ground surface.
There is currently an absence of site-specific geosphere data. However, two site
characterisation boreholes will be drilled in late 2011/early 2012 to allow site-specifc
data to be obtained, hence this first iteration of the Post-Closure Safety Assessment
(PCSA) will use data on the regional geology, hydrogeology and geochemical
conditions and extrapolate to the site.
1
FIG. A. Schematic Representation of a Borehole Site
2
Figure. B. Illustrative section through a Disposal Borehole
The report documents the post-closure radiological safety assessment for the
borehole disposal concept, with the purpose of identifying the safety of the
implementation of this concept in Ghana.
The PCSA has been undertaken using an approach that is consistent with best
international practice. Specifically, the approach developed by the Coordinated
Research Project of the International Atomic Energy Agency (IAEA) on Improving
Long Term Safety Assessment Methodologies for Near Surface Radioactive Waste
Disposal Facilities (the ISAM Safety Assessment Approach) has been used, with the
aim of ensuring that the assessment is undertaken and documented in a consistent,
logical and transparent manner. The ISAM Safety Assessment Approach consists of
the following key steps:
 the specification of the assessment context;
 the description of the disposal system;
 the development and justification of scenarios;
 the formulation and implementation of models; and
 the presentation and analysis of results.
3
Each of these steps is applied to the PCSA of the borehole disposal concept and the
application is described in this report.
The main report is supported by a series of appendices that provide detailed
information relating to specific aspects of the assessment study, namely:
 the approach used to identify scenarios and conceptual models for
consideration in the PCSA and the screening of associated features, events and
processes (in particular those associated with the borehole itself);
 the detailed models used to undertake the calculations of cement degradation
and the corrosion of stainless steel waste capsules and disposal containers in
the different environmental conditions considered;
 the assessment-level models and data used to calculate the impacts of disposals
to the borehole disposal concept; and
 the results of the associated calculations.
The PCSA has been developed so that it can serve as the primary post-closure safety
assessment for Ghana’s disposal site that lies within the envelope of conditions
assessed in this report.
4
CONTENTS
EXECUTIVE SUMMARY...................................................................................... 1
1
INTRODUCTION…………………………………………………………..9
1.1
BACKGROUND……………………………………………………………9
1.2
OBJECTIVE……………………………………………………………….10
1.3
SCOPE …………………………………………………………………….10
1.4
STRUCTURE………………………………………………………………11
2
SPECIFICATION OF ASSESSMENT CONTEXT……………………….13
2.1.
PURPOSE AND SCOPE…………………………………………………..13
2.2.
TARGET AUDIENCE……………………………………………………..13
2.3.
REGULATORY FRAMEWORK………………………………………….14
2.4.
ASSESSMENTEND POINTS……………………………………………...15
2.7
TIMEFRAMES……………………………………………………………..20
3
DESCRIPTION OF THE DISPOSAL SYSTEM .......................................... 22
3.1
NEAR FIELD……………………………………………………………….23
3.1.1 Inventory ....................................................................................................... 23
3.1.2
Engineering ............................................................................................. 27
3.1.3
Hydrology and Chemistry ....................................................................... 33
3.1.4
Safety Related Functions ......................................................................... 34
3.1.5
Uncertainties............................................................................................ 37
3.2
GEOSPHERE……………………………………...………………………..37
3.2.1
Structural Geology and Stratigraphy........................................................ 37
3.2.2 Seismicity ...................................................................................................... 39
5
3.2.3
Hydrogeology .......................................................................................... 40
3.2.4
Geochemistry .......................................................................................... 41
3.2.5
Natural Resources ................................................................................... 44
3.2.6
Safety Related Functions ......................................................................... 44
3.2.7
Uncertainties............................................................................................ 44
3.3
BIOSPHERE…………………………………….………………………….45
3.3.1
Topography ............................................................................................. 45
3.3.2
Climate .................................................................................................... 45
3.3.3
Surface Water Bodies .............................................................................. 46
3.3.4
Human Activity and Biota....................................................................... 46
3.3.5
Near-surface Lithostratigraphy ............................................................... 47
3.3.6
Safety Related Functions ......................................................................... 47
3.3.7
Uncertainties............................................................................................ 47
4
IDENTIFICATION AND DESCRIPTION OF SCENARIOS ............... 48
4.1
APPROACH…………………………………………………………………48
4.2
DESIGN SCENARIO………………………………………………………50
4.2.1
Description ........................................................................................... 50
4.2.2
FEP Screening ......................................................................................... 54
4.3
DEFECT SCENARIO………………………………………………………55
4.3.1
Description .............................................................................................. 55
4.3.2
FEP Screening ......................................................................................... 55
6
4.4
Unexpected Geological Characteristics Scenario…………………………...57
4.4.1
Changing Environmental Conditions scenario ........................................ 57
4.4.2
Borehole Disturbance Scenario ............................................................... 58
5
DEVELOPMENT AND IMPLEMENTATION OF MODELS ............. 59
5.3
APPROACH………………………………………………………………..59
5.2
CONCEPTUAL MODELS…………………………………………………60
5.2.1
Near Field ................................................................................................ 60
5.2.2
Geosphere ................................................................................................ 64
5.2.3
Biosphere................................................................................................. 65
5.3
MATHEMATICAL MODELS……………………………………………..67
5.3.1
Assessment Model................................................................................... 67
5.3.2
Supporting Models .................................................................................. 68
5.4
DATA…………………………………………………………………….....70
5.5
IMPLEMENTATION………………………………………………………70
6
PRESENTATION AND ANALYSIS OF RESULTS ............................ 72
6.1
RESULTS FOR THE REFERENCE CALCULATIONS…………………..74
6.1.1 Design Scenario ........................................................................................... 74
6.1.2
Defect Scenario ......................................................................................... 76
6.2
RESULTS FOR VARIANT CALCULATIONS…………………………...78
6.3
ANALYSIS OF UNCERTAINTIES………………………………………..81
6.4
BUILDING OF CONFIDENCE……………………………………………83
7
CONCLUSIONS ..................................................................................... 86
REFERENCES........................................................................................................... 88
7
APPENDIX A: SCREENING OF SOURCES .......................................................... 94
A.1
DECAY-STORAGE SCREENING………………………………………..95
A.2
ASSESSMENT SCREENING…………………………………………….98
APPENDIX B: APPROACH FOR CONCEPTUAL MODEL DEVELOPMENT . 101
B.1
NEAR-FIELD COMPONENTS………………………………………….101
GEOSPHERE COMPONENTS…………………………………………………. 102
B.3
BIOSPHERE COMPONENTS……………………………………………102
B.4
INTERACTIONS BETWEEN COMPONENTS ………………………..103
References for Appendix B .................................................................................. 103
APPENDIX C: ASSESSMENT MODEL ............................................................... 105
C.1
DESIGN SCENARIO…………………………………………………….105
C.1.1
Release Processes ................................................................................. 105
C.1.3
Exposure Mechanisms........................................................................... 116
C.2
DEFECT SCENARIO…………………………………………………….120
C.3
REPRESENTING NEAR-FIELD DEGRADATION……………………120
C.3.1
Physical Performance ............................................................................ 120
C.4.2
Chemical Performance .......................................................................... 122
APPENDIX D: ASSESSMENT DATA .................................................................. 124
D.1
INVENTORY AND RADIONUCLIDE DATA…………………………129
D.2
ELEMENT-DEPENDENT DATA………………………………………..133
D.3
NEAR-FIELD ELEMENT-INDEPENDENT DATA……………………142
D.4
GEOSPHERE ELEMENT-INDEPENDENT DATA……………………151
D.5
BIOSPHERE ELEMENT-INDEPENDENT DATA…………………….156
8
1
INTRODUCTION
1.1
BACKGROUND
The application of radioactive sources in medicine, research, industry, agricultural
and consumer products is a world-wide phenomenon. Ghana has disused sources
that need to be managed and disposed of carefully and in a safe and secure manner.
These sources contain different radionuclides in highly variable quantities. Some of
these sources have decayed to a level below which the source is no longer suitable
for its original purpose, in others the associated equipment has become obsolete,
worn out, or damaged, and in others the source has develop a leak and so is no
longer used. Even though these radioactive sources are referred to as ‘disused’ or
‘spent’1, the activities of some of them are still very high. Furthermore, despite their
predominately small physical size, radioactive sources can contain very high
activities, in Ghana for instance, the activities of disused radioactive sources ranges
from 7.4E+5 Bq to 6.85E+14 Bq. Therefore, if they are not managed properly,
radioactive sources can represent a significant hazard to human health and the
environment, which is evident from the number of accidents that have taken place
world-wide as a result of the mismanagement of such sources IAEA (2001). Ghana
has secure storage facility (Figure 1) which is considered as an adequate final
management option for sources containing quantities of short-lived radionuclides,
which decay to harmless levels within a few years. However, for most other sources
a suitable disposal option is required that will provide higher levels of isolation than
surface storage or near-surface facilities.
1
According to IAEA (2007) subtle differences can be noted between the terms ‘spent’ and ‘disused’. A disused
source differs from a spent source in that it may still be capable of performing its function, even though it is no
longer used for that purpose. To be consistent, the broader ‘disused’ term is used in this document.
9
Figure 1: Ghana’s Radioactive Waste Storage Facility for Disused Sources.
As Ghana does not have an extensive nuclear power programme that would require
the construction of a deep geological disposal facility, the disposal of disused
sources in narrow diameter (a few tens of centimetres) borehole facilities would
appear to provide a safe and cost effective disposal option for its disused sources.
1.2
OBJECTIVE
The objective of this report is to document the first iteration of a post-closure safety
assessment (PCSA) for the implementation of the borehole disposal concept for
disused sources in Ghana.
1.3 SCOPE
The focus of the work described in this report is the post-closure, radiological safety
assessment of the disposal of disused radioactive sources in Ghana. The report
considers exposure of humans due to natural processes and human intrusion, but
excludes intrusion that can be considered as deliberate (i.e. intrusion by a human
when the intruder knows that the facility is a radioactive waste disposal facility).
Consistent with (ICRP, 2000), the impact of deliberate human intrusion is considered
10
to be the responsibility of those intruding and is beyond the scope of the current
assessment, as are malicious acts that might arise from deliberate human intrusion.
A variety of borehole designs have been used for the disposal of radioactive waste
with differing depths (a few metres to several hundred metres) and diameters (a few
tens of centimetres to several metres) (see (IAEA, 2005) for details). The design
evaluated in the PCSA is based on the narrow diameter (0.26 m) design developed
under the AFRA project of the International Atomic Energy Agency (IAEA)
(NECSA, 2003) since this design has been developed specifically for the disposal of
disused radioactive sources and uses borehole drilling technology that is readily
available in Ghana. The design can accommodate disused sources of less than 110
mm in length and 15 mm in diameter.
It is recognised that, while radiological safety is of key importance, it is still only
part of a broader range of issues that need to be considered in a safety case such as
planning, financial, economic and social issues, and non-radiological safety (IAEA,
2002). However, these issues are not specifically covered in this report. They need
to be considered as part of the wider safety case documentation that should be
developed to support the licence application for the construction and operation of the
borehole.
1.4 STRUCTURE
The PCSA has been undertaken using an approach that is consistent with
international best practice, as embodied in the draft safety standards on the safety
case and safety assessment for radioactive waste disposal from the International
Atomic Energy Agency (IAEA) (IAEA, 2010) and the recommendations of the
IAEA programme for the Improvement of Safety Assessment Methodologies
(ISAM)(IAEA, 2004a, b) (Figure2). This has ensured that the assessment has been
undertaken and documented in a consistent, logical and transparent manner. The
approach consists of the following key steps:
 the specification of the assessment context;
 the description of the disposal system;
 the identification and description of scenarios;
 the development and implementation of models; and
 the presentation and analysis of results.
These steps are presented in Sections 2 to 6 with the overall conclusions being
presented in Section 7
11
First Iteration of PCSA
Specify Assessment
Context
Describe Disposal System
Identify and Describe
Scenarios
Develop and Implement
Models
Run Analysis
Compare with Acceptance
Criteria and other Safety
and Performance Indicators
Present and Analyse
Results
Initial PCSA input to
Safety Case
Figure 2: The Safety Assessment Approach Used
12
2
SPECIFICATION OF ASSESSMENT CONTEXT
The assessment context defines the scope and content of the safety assessment.
Specifically, it specifies the assessment’s:
 purpose and scope (Section 2.1);
 target audience (Section 2.2);
 regulatory framework (Section 2.3);
 assessment end-points (Section 2.4);
 treatment of uncertainties(Section 2.5);
 building of confidence(Section 2.6); and
 timeframes (Section 2.7).
2.1. PURPOSE AND SCOPE
The PCSA has three main purposes.

To produce the first iteration of a site-specific post-closure safety assessment
that can be used in the development of a suitable borehole disposal facility
(BDF) for Ghana taking into account the inventory to be disposed and the site
characteristics.

To identify the key parameters that needs to be characterised at the proposed
site.

To demonstrate and build confidence in the use of narrow diameter boreholes
as a safe disposal concept for disused radioactive sources of less than 110
mm in length and 15 mm in diameter.
The PCSA’s scope is the assessment of the post-closure (i.e. once the waste has been
emplaced and the borehole backfilled and closed) radiological impacts on humans
arising from the disposal of disused radioactive sources at least 50m below the
ground surface in a narrow diameter borehole.
2.2. TARGET AUDIENCE
This report is a technical report and as such is written primarily for a technical
audience whose prime interest is in the regulation and implementation of safe
13
radioactive waste disposal2.
The main technical audiences are:



the implementer, i.e. the National Radioactive Waste Management Centre
(NRWMC) of the National Nuclear Research Institute and its support
scientists, which is developing the safety case and associated PCSA for the
BDF in Ghana;
the regulator, i.e. the Radiation Protection Board and its support scientists,
which has a direct responsibility to decide whether to grant a licence to
construct, operate and close a BDF in Ghana; and
any international peer reviewers (e.g. IAEA appointed experts) that might
review the work.
It is recognised that there is a range of other audiences that could be interested in the
borehole disposal of disused sources (for example the media, politicians, and the
public). However, given its technical focus, this report is not specifically aimed at
these audiences. It is recognised that additional data will have to be developed that
is tailored to the specific interests and needs of these other audiences.
2.3. REGULATORY FRAMEWORK
As the assessment is related to the disposal of disused sealed sources in Ghana, there
is the need to use Ghana-specific legislation. The 1993 Radiation Protection
Statutory Instrument established the Radiation Protection Board and the regulatory
infrastructure. Ghana-specific legislation for radioactive waste disposal is in the
preparatory stage, so the recommendations of the IAEA safety guide for the borehole
disposal facilities for radioactive waste (IAEA, 2009) are adopted. This safety guide
provides post-closure protection objectives and criteria which in turn are based on
the recommendations of the IAEA (1996, 2006) and ICRP (2000).
Consistent with IAEA (2009), this first iteration of the PCSA adopts an individual
effective dose constraint of 0.3 mSvy-1 for adult3 members of the public for all
potential future exposures other than those arising from human intrusion. In future
iterations of the PCSA, infants and children can be considered as well. For exposures
arising from human intrusion, ICRP (2000) recommends that, if human intrusion is
expected to lead to an annual dose of less than about 10 mSvy-1 to those living
around the site, efforts to reduce the probability of human intrusion or to limit its
2
The report assumes that the reader is familiar with the technical terms used in safety assessment. Key technical
terms are defined in IAEA (2007).
3
Doses to children and infants could also be calculated, especially if there was a need to demonstrate
consideration of a wide
range of calculation end points. However, various post-closure assessment studies, such as IAEA (2003), Prӧhl
et. al. 2004, have demonstrated that the differences between adult, child and infant doses are usually less than a
factor of two. Therefore, for the purposes of the PCSA, consideration will be limited to adult doses as an
indicator of impacts.
14
consequences are not likely to be justifiable. If human intrusion is expected to lead to
an annual dose of more than about 100 mSv y-1 to those living around the site, then it
is almost always justifiable to make reasonable efforts at the stage of development of
the facility to reduce the probability of human intrusion or to limit its consequences.
Radiological impacts on non-human biota are not considered in this report since it is
assumed that if individual humans are shown to be adequately protected, then nonhuman biota will also be protected, at least at the species level (ICRP 1991). The
basis of this assumption is currently being investigated by various international
organisations such as the International Commission on Radiological Protection
(ICRP), the IAEA and the European Commission. However, in the absence of any,
as yet, clear consensus and guidance on the assessment of radiological impacts on
non-human biota, the recommendations of ICRP Publication 60 (ICRP 1991) are
adopted.
Non-radiological impacts on both humans and non-human biota, which might arise
from the content of chemically or biologically toxic materials in the waste (for
example beryllium in some Am-241 sources) or engineered barrier materials, are
considered to be beyond the scope of this first iteration of the Ghana-Specific PCSA
given its emphasis on radiological impacts. Future iterations could be extended to
consider these impacts.
2.4. ASSESSMENT END POINTS
Assessment end points allow potential impact and the performance of the disposal
facility to be evaluated. They can be categorised as either safety indicators or
performance indicators (Marivoet et al., 2008). A safety indicator:




provides a statement on the safety of the whole disposal system;
provides a contaminant-specific or an integrated measure describing the
effects of the whole radionuclide spectrum;
is a calculable time-dependent parameter; and
allows comparison with safety-related reference values.
In contrast, a performance indicator:





provides a statement on the performance of the whole system, a subsystem or
a single barrier;
provides a contaminant-specific or integral measure;
is a calculable, time-dependent or absolute parameter;
allows comparison between different options or with technical criteria; and
illustrates the functioning of the disposal system.
The following safety indicators are considered for the PCSA:


radiation dose to adults; and
environmental concentrations of radionuclides.
For the performance indicators, the following are evaluated:
15



the amount of radionuclides in various regions (borehole, geosphere and
biosphere) of the disposal system;
the fluxes of radionuclides at various points in the disposal system; and
the radiotoxicity of the waste.
The long-term assessment of impacts, e.g. calculated dose, are not absolute values
and they must be seen as estimates since the reliability of quantitative predictions
diminishes with increasing time (IAEA, 2006).
2.5 TREATMENT OF UNCERTAINTIES
The treatment of uncertainty is an important aspect of any assessment of the safety of
a radioactive waste disposal facility.
The following three broad categories are used by many organisations to structure
their analysis of uncertainties in post closure safety assessments (Marivoet et al.,
2008):

Future or scenario uncertainty: uncertainty in the evolution of the disposal
system and human behaviour over the timescales of interest;

Model uncertainty: uncertainty in the conceptual, mathematical and
computer models used to simulate the behaviour of the repository system
(e.g., due to approximations used to represent the system); and

Data uncertainty: uncertainty in the data and parameters used as inputs in
the modelling (e.g., due to incomplete site-specific data, and parameter
estimation errors from interpretation of test results).
Uncertainties are accounted for in the current safety assessment through:



the assessment of a range of scenarios, models and data with deterministic
calculation cases;
the adoption of conservative scenarios, models and data, where appropriate;
and
the adoption of a stylised approach for the representation of future human
actions and
biosphere evolution.
2.5.1 Range of Scenarios, Models and Data
In the PCSA, a range of scenarios that describe the potential evolution of the system
have been used to address the uncertainty in the future evolution of the site and
human behaviour (Section 4). The process of identifying and justifying scenarios,
ensures that scenarios are defined to investigate the consequences of key
uncertainties that are identified. Some future uncertainties can be investigated in the
same way as data uncertainties and can be represented by varying parameter values.
16
Various Features/Events/Processes (FEPs) are used in the model development
process to identify conceptual and mathematical models (Section 5). The availability
of a computer code, AMBER, that is capable of representing different
conceptualisations and mathematical descriptions of the system allows alternative
conceptual representations of the system to be developed to address key conceptual
and mathematical model uncertainties. Here again, some model uncertainties can be
represented by varying parameter values.
The multiple deterministic calculations in which alternative sets of parameter values,
which provide a self-consistent representation of the system, are adopted to analyse
data uncertainties (Section 6).The impact of specific uncertainties or uncertainty
combinations is achieved by comparing the results of variant cases to those of the
Reference Case and the discrepancies explored. In future iterations of the PCSA,
probabilistic calculations could be used to complement the deterministic
calculations.
2.5.2 Conservative Scenarios, Models and Data
Different assumptions relating to scenarios, models or data have to be made during
the assessment process. These assumptions can be categorised as ‘realistic4’ or
‘conservative5’, although, as noted in IAEA (2006b), care needs to be taken when
using such terms. The key is to ensure that the nature of each major assumption used
in the assessment is considered and documented, and that the potential implications
are understood (see Section 6).
In this PCSA, conservative assumptions have been adopted where there are high
levels of uncertainty associated with the scenarios, processes and/or data being
evaluated. For scenarios, processes and/or data that are understood and can be
justified on the basis of the results of site investigation and/or research, realistic
assumptions have been used.
2.5.3 Stylised Approach6
4Realism
is defined as “the representation of an element of the system (scenario, model or data), made in light of
the current state of system knowledge and associated uncertainties, such that the safety assessment incorporates
all that is known about the element under consideration and leads to an estimate of the expected performance of
the system attributable to that element” (IAEA 2006b).
5
Conservatism is defined as “the conscious decision, made in light of the current state of system knowledge and
associated uncertainties, to represent an element of the system (scenario, model or data) such that it provides an
under-estimation of system performance attributable to that element and thereby an over-estimate of the
associated radiological impact (i.e., dose or risk)” (IAEA 2006b).
6A
stylised representation of the biosphere, and human habits and behaviour is a representation that has been
simplified to reduce the natural complexity to a level consistent with the objectives of the analysis using
17
It is unrealistic to predict human habits and behaviour over the timescale of potential
relevance to the BDF. Further, major changes to the surface and near-surface
environment are also likely as a result of natural changes such as erosion or as a
result of future human actions. Thus, in order to estimate the potential future impacts
of the BDF, a ‘reference’ biosphere approach has been adopted, consistent with the
recommendations of the international BIOMASS and BIOCLIM programmes (IAEA
2003, BIOCLIM 2004). In this approach, stylised representations of the biosphere
are used to allow illustrative estimates of impact to be made. Each stylised biosphere
acts as a ‘measuring instrument’ for evaluating the safety and performance indicators
identified in Section 2.4.
2.6 BUILDING OF CONFIDENCE
Through discussions within various international bodies such as the Nuclear Energy
Agency, (NEA 2004a, b) and the IAEA (IAEA 2003, 2004), it is becoming
recognised that building confidence in the long-term safety of a radioactive waste
disposal facility is an increasingly important issue. To undertake a safety assessment
and present the results is not sufficient. Confidence needs to be built in the safety
assessment and its results.
Confidence building can be achieved by (NEA,1999a; IAEA, 1999):
 the use of a systematic assessment methodology that allows the assessment to
be undertaken using a well-structured, transparent and traceable manner;
 the use of an iterative approach that allows the results of previous
assessments to be used to inform the current assessment;
 the use of a range of strategies to identify and manage the various
uncertainties associated with the assessment;
 the demonstration that the repository system will maintain its integrity and
reliability under extreme conditions (i.e. the system is robust);
 the use of multiple lines of evidence to support key findings;
 the application of a quality management system to the assessment;
 the peer review of the assessment and its results; and
 the comparison of the repository system with natural systems that have
evolved over
relevant timescales.
Confidence of stakeholders in a PCSA can be established at two levels (IAEA,
2003).
The first level involves establishing confidence within each stage of the assessment
process (i.e. assessment context, system description, development and justification of
scenarios, formulation and implementation of models and associated data, analysis of
the results, and review and modification).
assumptions that are intended to be plausible and internally consistent but that will tend to err on the side of
conservatism.
18
The second level involves gaining overall confidence in the PCSA and associated
implications for further data gathering, assessment and design optimisation.
Various measures and attributes that can be used to develop confidence in the
assessment at these two levels are summarised in Table 1.
Table 1: Confidence Building Measures and Attributes
Confidence in each Stage of the Assessment Process
Assessment
Stage
Confidence Building Measures and Attributes
Assessment
Context
 Demonstration of understanding of the key
components of the assessment context.
System
Description
 Demonstration of sufficient understanding of
engineered and natural aspects of the borehole
disposal system (near field, geosphere and
biosphere) and associated uncertainties.
 Linkage to waste and site characterisation, and
borehole design.
Scenarios
 The set of scenarios is adequately comprehensive
and is developed in a systematic, transparent and
traceable way.
 The approach used to exclude or include
scenarios is justified and well documented.
 Scenarios are consistent with the waste and site
characterisation, and borehole design.
Models
Data
and
Analysis
Results
of
 The conceptual models and associated data are
consistent with the assessment context, borehole
disposal system, and scenarios.
 The software tools used adequately solve the
problems under consideration.
 Alternative models, codes, data and approaches
are considered.
 Models are consistent with the assessment
context, waste and site characterisation, and
borehole design.




Key assumptions are documented and justified.
Results are reasonable and understandable.
Uncertainties are adequately addressed.
Conformity with regulatory requirements and
recommendations is analysed.
Confidence in the
Overall Safety of the
BDF
 Use of a
systematic
approach
consistent with
international
practice and
recommendations.
 Adequate
understanding of
the borehole
disposal system
and its
uncertainties.
 Use of multiple
safety and
performance
indicators.
 Clear presentation
of the assessment
and its results.
 Application of a
quality
management
system.
 Peer review of the
assessment.
 Involvement of
stakeholders in the
development of
the assessment.
19
Confidence in each Stage of the Assessment Process
Assessment
Stage
Review and
Modification
2.7
Confidence Building Measures and Attributes
Confidence in the
Overall Safety of the
BDF
 Modifications are implemented in an organized
and well-documented manner.
TIMEFRAMES
Table 2 summarises the timeframes for the various activities associated with the
construction, operation, closure and subsequent release of the borehole from
institutional control. It is assumed that following construction of the borehole, waste
is disposed for a maximum period of one year since the volume of waste packages to
be disposed is small (less than 0.2 m3) and, from an operational (and post-closure
safety) perspective, it is best for this to be disposed over a relatively short period of
time. It is assumed that following the disposal of the disused sources, the site is
closed immediately and the institutional control period starts. During this period (i.e.
50 years), surveillance of the site might be undertaken for the purpose of public
assurance (active institutional control), and local/national government records,
planning authority restrictions maintained to prevent unauthorised use of the land
and inadvertent human intrusion (passive institutional control). During the
institutional control period, it is assumed that members of the public do not have
access to the land in the immediate vicinity of the borehole and that inadvertent
human intrusion into the facility does not occur.
It is worth noting that, in the absence of guidance in legislation, this initial
assessment uses an institutional control period of 50 years as the reference duration
but sensitivity analysis is undertaken using alternative durations ranging from 30 to
100 years.
20
TABLE 2. TIMEFRAMES FOR THE VARIOUS ACTIVITIES ASSOCIATED
WITH THE CONSTRUCTION, OPERATION, CLOSURE AND SUBSEQUENT
RELEASE OF THE BOREHOLE FROM INSTITUTIONAL CONTROL
Activity
Borehole Construction and Waste Emplacement
Site Closure
Timeframe
One year
Immediately following the waste
disposal operation
Institutional Control Period (e.g. surveillance,
local/national government records, planning authority
restrictions, site marked on official maps)
50 years
No control (neither active nor passive) – all
records/knowledge conservatively assumed to be lost
From 50 years onwards
In terms of the cut off time for calculations, the regulatory framework adopted for
the assessment does not impose any explicit limit on the timescale for assessment.
Therefore, calculations presented in the PCSA are undertaken out to a time when it
can be demonstrated that the peak value of the primary safety indicator (dose) has
been passed for the radionuclide and disposal system of interest. It is important to
recognise that uncertainties associated with these estimates will increase as the
timescales become longer.
21
3
DESCRIPTION OF THE DISPOSAL SYSTEM
Together with the assessment context, the disposal system description provides the
necessary basis to develop a well-justified set of exposure scenarios (Section 4). The
proposed location of the BDF is on Ghana Atomic Energy Commission (GAEC)
land on the Accra plains (Figure 3).
Figure 3: The GAEC Site
The disposal system can be divided into:
 the near field - the waste, the disposal zone, the engineered barriers of the
borehole plus the disturbed zone of the natural barriers that surround the
borehole;

the geosphere - the rock and unconsolidated material that lies between the near
field and the biosphere. It can consist of both the unsaturated or vadose zone
(which is above the groundwater table) and the saturated zone (which is below
the groundwater table); and

the biosphere - the physical media (atmosphere, soil, sediments and surface
waters) and the living organisms (including humans) that interact with them.
These descriptions are provided in Section 3.1 to 3.3.
22
3.1
NEAR FIELD
In the case of PCSA, there is a single disposal borehole and that the design assessed
is based on the narrow diameter borehole design developed under the IAEA’s AFRA
project (NECSA, 2003) It is assumed that the disposal zone in the borehole is 56.5 m
from the ground surface thereby significantly reducing the probability of the waste
being disturbed by human intrusion or other disruptive events and processes (IAEA,
2005). The disposal zone extends down to 100 m.
3.1.1 Inventory
In Ghana, radioactive waste is generated mainly from research, medical and
industrial applications. The current inventory includes a range of disused sources that
can neither be repatriated nor decay stored. It is this inventory of disused sources
that is to be disposed in the BDF.
A national inventory of wastes is being compiled by NRWMC and currently lists
disused sealed and unsealed sources. Supplementing this national inventory of
disused sources is an inventory of sources-in-use that is compiled by the Radiation
Protection Institute (RPI).
The list of the radionuclides found so far in disused sources in Ghana is given in
Table 3.
TABLE 3.Ghana’s Inventory of Disused Sources
Radionuclide
Total Initial Application
Activity (Bq)
Form
Quantity
Cs-137
5.66x1012
Level Gauges
Sealed
30
Co-60
1.75x106
Non Destructive Sealed
Testing (NDT)
2
Cs-137/Co-60
4.09x106/
4.90x105
Not specified
Sealed
2
Cs-137/Am-241
3.70x1011/
1.85x1012
Not specified
Sealed
3
Cs-137/Am241:Be
3.00x107/
1.80x109
Nuclear Gauges
Sealed
1
Am-241
3.50x107
Smoke Detectors
Sealed
105
23
Sr-90
1.25x1010
Thickness Gauges Sealed
33
Ir-192
2.26x1012
NDT
Sealed
1
Cd-109
6.66x108
Research
Sealed
6
Am-241
1.67x109
Nuclear Gauge
Sealed
1
I-131
6.21x109
Not specified
Unsealed
2
Cf-252
2.22x1010
Not specified
Sealed
2
Ra-226
7.03x109
Not specified
Sealed
19
H-3 (1)
3.70x107
Nuclear Gauge
Unsealed
(Liquid)
218 litres
C-14 (2)
Originally
Not specified
contained
2.60x107 but
now empty
Unsealed
(Gas)
7 empty
cylinders
DISUESD HIGH DOSE SOURCES
Radionuclide
Total Initial Application
Activity (Bq)
Co-60
2.78x1014
Gamma
research
Co-60
1.85x1014
Co-60
2.22x1014
Form
Quantity
Cell- Sealed
1
Teletherapy
Sealed
1
Food Irradiator
Sealed
1
UNCHARACTERISED SOURCES
I-129
4.25x1010
Not specified
Sealed
(assumed
to be in
solid
form)
1
Fe-59
2.22x1010
Not specified
Sealed
2
Co-57
1.11x108
Not specified
Sealed
3
24
Zn-65
3.70x108
Not specified
Sealed
1
Sr-89
4.77x109
Not specified
Sealed
1
Tl-204
7.40x105
Not specified
Sealed
2
P-32
1.18x109
Research
Unsealed
4
S-35 (3)
9.25x10-6/ml
Not specified
Unsealed
5
Ca-45
1.85x108
Not specified
Unsealed
3
Na-22
3.7x106
Not specified
Unsealed
4
In-113m
2.22x109
Not specified
Unsealed
12
Notes
1. Not suitable for disposal to the BDF due to the waste being liquid. The
volume of solidified waste would be too large for a single BDF. Therefore,
this waste is excluded from consideration in the current PCSA.
2. Containers are empty so there is no inventory to be disposed.
3. No data currently available on the volume of waste and so excluded from
consideration in the current PCSA.
A number of sources in Table 3 contain radionuclides with half lives of much less
than a year and so could potentially be decay stored rather than disposed in the BDF.
In order to identify suitable sources for decay storage, a spreadsheet has been
developed to allow the calculation of doses associated with direct exposure via
ingestion, inhalation and external irradiation to a source (see Appendix A.1). The
calculations indicate that the P-32, Ca-45, Fe-59, Sr-89, In-113m, I-131 and Ir-192
sources can all be decay stored and do not need to be considered for disposal in the
BDF.
Of the remaining sources that are to be disposed in the BDF, it is possible to
undertake a further screening calculation to identify those sources which contain
radionuclides that, due to their half-life, maximum activity, and/or radiotoxicity, will
not result in significant post-closure impacts and so do not need to be assessed in
detail (see Appendix A.2). Doses associated with direct exposure via ingestion,
inhalation and external irradiation to a source following a 50 year institutional
control period are calculated for each type of source identified in Table 3 and not
suitable for decay storage. A dose constraint of 0.3 mSv y-1 is applied (Section 2.3).
This screening process results in the identification of the sources listed in Table 4 for
more detailed consideration in the PCSA. The screening calculations show that the
sources containing Na-22, Co-57, Zn-65, Cd-109 and Tl-204 can be safely disposed
in the BDF and do not need to be assessed in more detailed.
25
Table 4 provides details on the assumed dimensions of the sources that require more
detailed consideration. At present, the dimensions of the sources have not been
measured. Therefore for the purposes of the PCSA, dimensions have generally been
derived using information from Section 6 and/or Table 2 of Appendix III of IAEA
(2007). Values have been chosen from the upper end of the ranges quoted in IAEA
(2007).
Table 4: Sources Screened in for Detailed Consideration in the PCSA
Radionuclide
Nature of Source
Number
of
sources
Total
Initial
Activity
(Bq)
Dimensions (mm)
Note
Co-60
NDT
2
1.75E+06
A
Co-60
1
2.78E+14
Co-60
Gamma Cellresearch
Teletherapy
1
1.85E+14
Co-60
Food Irradiator
1
2.22E+14
Sr-90
Thickness gauges
33
1.25E+10
Diameter: 7
Length: 15
Diameter: 8
Length: 20
Diameter: 20
Length: 30
Diameter: 11
Length: 450
Diameter: 15
Length: 10
I-129
Currently
unknown
Level gauges
1
4.25E+10
F
30
5.66E+12
3
3.7E+11
Diameter: 15
Length: 15
Diameter: 12
Length: 15
Diameter: 12
Length: 15
1
3.00E+7
Diameter: 12
Length: 15
G
19
7.03E+09
G
3
1.85E+12
Diameter: 12
Length: 15
Diameter: 12
Length: 15
1
1.80E+09
Diameter: 12
Length: 15
G
105
3.50E+07
Diameter: 15
Length: 10
H
Cs-137
Cs- 137
Cs- 137
Ra-226
Am-241
Am-241
Am-241
Currently
unknown (also
contains Am-241)
Nuclear gauge
(also contains
Am-241)
Currently
unknown
Currently
unknown (also
contains Cs-137)
Nuclear gauge
(also contains
Cs-137)
Smoke detectors
B
C
D
E
G
G
G
26
Am-241
Nuclear gauge
1
1.67E+09
Cf-252
Currently
unknown
2
2.22E+10
Diameter: 12
Length: 15
Diameter: 20
Length: 30
G
I
Note:
A: Assumed to be an industrial gamma radiography source.
B: Assumed to be a high activity gamma source.
C: Typical dimensions of a source for teletherapy.
D:Typical dimensions of a source for food irradiation.
E: Assumed to be a low energy fixed industrial guage source.
F:No data given in IAEA (2007) on dimensions of source, therefore adopt assumed
dimensions for the purposes of the current PCSA.
G: as assumed to be a high energy gamma industrial gauge source.
H: No data given in IAEA (2007) on dimensions of source, therefore adopt assumed
dimensions for the purposes of the current PCSA.
I: Assume to be neutron industrial gauging source.
3.1.2
Engineering
Based on the narrow diameter design developed under the IAEA’s AFRA project
(NECSA, 2003) the reference design for the near field comprises a series of
engineered components which are described below, illustrated in Figures 4 to 6, and
summarised in Table 5.
Figure 4: Schematic Representation of the Borehole Site (Van Blerk, 2000)
27
Figure 5: Cross-section through the Disposal Borehole for the Reference Design
TABLE 5. NEAR-FIELD COMPONENTS FOR THE REFERENCE DESIGN
Near-field
Component
Source and
its container
Capsule
Containment
barrier
Disposal
container
Disposal
zone backfill
Disposal
zone plug
Description
Source and its container within which the source material is sealed
Standard stainless steel (Type 304) capsule containing the source container
Space between the capsule and the disposal container is backfilled with sulphateresistant cement grout
Type 316 L stainless steel
Sulphate-resistant cement grout used to separate disposal containers in vertical
dimension from one another, and in the horizontal dimension from the borehole
casing
Sulphate-resistant cement grout plug at base of borehole
28
Near-field
Component
Casing
Description
Disturbed
zone backfill
High-density polyethylene (HDPE) casing emplaced at time of drilling. Top sections
withdrawn at closure of borehole down to 1 m of the disposal zone
Sulphate-resistant cement grout used to fill the gap between the casing and the host
rock and any voids/cracks in the host rock immediately adjacent to the borehole
Closure zone
backfill
Assume that the first 5 m from the ground surface is native soil/crushed rock and the
remainder is sulphate-resistant cement grout
Waste Package
The waste package used for the disposal of disused radioactive sources in the
borehole disposal concept comprises the following components (see Figure5). .
 The source and its container - the radioactive source material and its
container. The dimensions of the capsule (Table 6) limit the source and its
container to be less than 110 mm in length and 15 mm in diameter if the
small capsule is used, and 121 mm in length and 40 mm in diameter if the
large capsule is used.
 The capsule –is a standard stainless steel capsule (Type 304)7. The disused
source and its associated container are emplaced in the capsule and sealed.
No backfill material is used, which means that apart from the disused source
and its container, the capsule is empty. The dimensions for the small and
large versions of the capsule are presented in Table 6.
 The containment barrier –a backfill, comprising sulphate-resistant cement
grout, filling the void between the capsule and the disposal container. The
dimensions for the containment barrier are presented in Table 6.
 The disposal container - is manufactured from Type 316 L stainless steel
with the reference dimensions given in Table 6. As shown in Figure5, the
container is equipped with a lifting ring to facilitate waste emplacement in
the borehole. There are also three centralisers that help to ensure that the
container is emplaced centrally and vertically. The centralisers are thin (<10
mm) and do not inhibit the flow of cement grout past the top of the disposal
container.
7
Stainless steels are chromiun-containing steels where the Cr provides resistance to corrosion through the
formation of a protective (”passive”) Cr(III) oxide or hydroxide film. There are various classes of stainless steel,
a common class being the so-called 300-series austenitic alloys. Two of these alloys have been selected for the
waste capsule and disposal container. Type 304 stainless steel has the nominal composition 18-20 wt.%Cr, 810.5 wt.%Ni, 1 wt.%Si, 2 wt.%Mn, 0.08 wt.%C, 0.045 wt.%P, and 0.03 wt.%S. Type 316L stainless steel has
the nominal composition 16-18 wt.%Cr, 10-14 wt.%Ni, 1 wt.%Si, 2 wt.%Mn, 0.03 wt.%C, 0.045 wt.%P, 0.03
wt.%S, and 2-3 wt.% Mo, where the addition of Mo improves the resistance to localised corrosion and the
reduced C content improves resistance to intergranular attack.
29
TABLE 6. DIMENSIONS OF THE CAPSULE, CONTAINMENT BARRIER AND
DISPOSAL CONTAINER FOR THE REFERENCE DESIGN
Waste Package
Component
Capsule
Containment
Barrier
Length
(mm)
S
L
S
110
121
186
Inside
Diameter
(mm)
15
40
21
L
171
48
Outside
Diameter (mm)
Thickness1
(mm)
21
48
102
3
4
41
102
27
Disposal Container S,L
250
103
115
6
1 As used here thickness of the capsule and disposal containers as well as the thickness of
the containment barrier
Using the above dimensions, it can be calculated that a total of 43 waste packages
are required to dispose the inventory of disused sources in the BDF, assuming that
each source type is disposed in separate capsules/disposal containers (Table 7).
Table 7.
No. of
No. Of Diameter Length
Radionuclide Sources
(mm)
(mm) Capsules
Comments
Na-22
4
15
15
1
Dimensions were assumed
Co-57
3
15
15
1
Dimensions were assumed
Co-60
2
7
15
1
See Table 4 for Dimensions
Co-60
2
12
15
1
See Table 4 for Dimensions
Co-60
1
15
15
1
Co-60
1
20
30
1
Dimensions were assumed
See Table 4 for Dimensions
(need large capsule)
Co-60
1
11
450
5
See Table 4 for Dimensions
Zn-65
1
15
15
1
Dimensions were assumed
Sr-90
33
15
10
4
Cd-109
6
40
15
1
See Table 4 for Dimensions
Typical Dimensions for low
gamma analytical sources
(needs large capsule)
I-129
1
11
9
1
Cs-137
30
12
15
5
Cs-137
2
12
15
A
Dimensions were assumed
See Table 4 for Dimensions
See Table 4 for Dimensions
30
See Table 4 for Dimensions
Cs-137
3
12
15
1
Cs-137
1
12
15
1
Tl-204
2
15
15
1
Ra-226
19
12
15
4
Am-241
3
12
15
B
Am-241
1
12
15
C
Am-241
105
15
10
11
Am-241
1
12
15
1
See Table 4 for Dimensions
Dimensions were assumed
See Table 4 for Dimensions
See Table 4 for Dimensions
See Table 4 for Dimensions
See Table 4 for Dimensions
See Table 4 for Dimensions
See Table 4 for Dimensions
(need large capsule)
Cf-252
2
20
30
1
Note:
A: Joint source with Co-60
B: Joint source with Cs-137
C: Joint source with Cs-137
Disposal Borehole
The disposal borehole is 260 mm in diameter and is drilled to a depth of about
100 m. The borehole is fitted with a high-density polyethylene (HDPE) casing. The
inner and outer diameters of the casing are 140 mm and 160 mm, respectively,
giving a casing thickness of 10 mm.
Three distinct zones can be defined in the disposal borehole (see Figure 6).
31
Native Soil/
Crushed Roc k
Closure Zone
Bac kfill
Closure Zone
Anti-intrusion
barrier
Casing
Disturbed Zone
Bac kfill
Disposal Zone
Bac kfill
Waste Pac kages
Disposal Zone
Disposal Zone
Plug
Figure 6: Illustration of the Borehole Zones for the Reference Design
The disposal zone - the zone inside the casing in which the waste packages are
disposed. The base of the disposal zone is 99.5 m from the ground surface. A
0.5 m thick ‘plug’ of backfill slurry is emplaced at the base of the borehole.
The borehole backfill slurry is assumed to be sulphate-resistant cement grout.
Once the plug material is set, the waste packages are lowered into the borehole,
one at a time. After the emplacement of each waste package, backfill material
is poured over the waste packages to fill the 12.5 mm thick void between the
waste package and the casing wall, as well as a volume on top of the waste
package. The layer of backfill on top of the waste package should be 750 mm
deep. Together with the waste package, this constitutes a pitch height of 1 m
per waste package. Given that there are 43 waste packages to be disposed, the
total thickness of the disposal zone is 43.5 m. The closure zone: the zone
between the disposal zone and the ground surface. Once the waste packages
have been emplaced in the borehole, the casing in the closure zone is
withdrawn from the borehole from a depth 1 m above the disposal zone. This
removes a potential fast transit pathway to and from the disposal zone which
might arise once the casing has degraded. An anti-intrusion barrier (for
example a metallic ‘drill deflector’) is placed above the disposal zone in order
32
to deter/prevent human intrusion. The closure zone is then backfilled to a depth
5 m below the ground surface with the same backfill material used for the
disposal zone. The final 5 m of the closure zone is then backfilled with native
soil and/or crushed rock to the ground surface. The total depth of the closure
zone is 50 m, which is an appreciable depth design to lessen the likelihood of
human intrusion and to limit the type of intrusion that might occur.
The disturbed zone: the zone between the casing and the wall of the borehole. Voids
and cracks in the host geology immediately adjacent to the borehole are
assumed to be grouted and sealed during the drilling process with the same
slurry used for the backfilling of the disposal and closure zones. In addition, an
average gap of 50 mm between the casing and the borehole wall is backfilled
with the slurry using a pressure grouting technique (NECSA, 2004). As shown
in Fig. 3, the casing is fitted with centralisers to ensure that the casing is in the
middle of the borehole. These centralisers are made of thin mild steel plates
inserted vertically to ensure that they do not hamper the flow of the backfill
slurry.
The design of the borehole disposal concept is a final disposal concept that is not
designed to facilitate the retrieval of waste packages once disposed since, once each
waste package has been lowered into the borehole; it is backfilled into the borehole
with sulphate-resistant cement grout. Following the emplacement of the final waste
package, the closure zone above the waste package is also backfilled with sulphateresistant cement grout. This greatly reduces the possibility of sabotage or theft of the
disposed disused radioactive sources.
3.1.3
Hydrology and Chemistry
Geochemical conditions in the borehole will be determined by the interaction of the
borehole engineering and the host groundwater. The geochemical characteristics of
the host groundwater are discussed in Section 3.2. Their impact on near-field
geochemistry is considered in Section 5.2.
33
3.1.4
Safety Related Functions
The post-closure safety related functions of the near field are summarised in Table 8.
TABLE 8. POST-CLOSURE SAFETY RELATED FUNCTIONS FOR THE NEAR-FIELD COMPONENTS
Near-field
Component
Source and its
container
Capsule
Post-closure Safety Related Functions
 No safety function since it is assumed that the source container has failed prior to
disposal
 Until breached, isolates source from water, animals and humans
 Until breached, prevents escape of gas from source
 Once breached, limits release of radionuclides available for release from the capsule
until it has been corroded
Containment barrier  Physical barrier – can inhibit disruption of the disused source by surface erosion,
human intrusion, and biotic intrusion
 Physical barrier – once the disposal container has been breached, can limit flow of water
around the capsule due to low permeability
 Physical barrier – once the capsule has been breached, can act as low permeability
barrier to the migration of radionuclides from the borehole in liquid and gaseous phases
 Cement can passivate corrosion of stainless steel capsule and reduce chloride levels in
water through formation of calcium chloride
 Chemical barrier - once the capsule has been breached, can act as sorption barrier for
radionuclides released
 Chemical barrier - once the capsule has been breached, can act to regulate the
availability of radionuclides for release into water through its impact on the solubility of
radionuclides
34
Near-field
Component
Disposal container
Disposal zone
backfill
Disposal zone plug
Casing
Post-closure Safety Related Functions
 Until breached, isolates source container, capsule and containment barrier from water,
animals and humans
 Once it and the capsule are both breached, the disposal container can limit the fraction
of radionuclides available for release into the borehole until the entire container has
been corroded
 Physical barrier – can inhibit disruption of the disused source by surface erosion, human
intrusion, and biotic intrusion
 Physical barrier – can limit the flow of water around the disposal container due to low
permeability
 Physical barrier – once the disposal container and capsule have been breached, can act
as low permeability barrier to the migration of radionuclides from the borehole in liquid
and gaseous phases
 Cement can passivate corrosion of stainless steel capsule and reduce chloride levels in
water through formation of calcium chloride
 Chemical barrier - once the disposal container and capsule have been breached, can act
as sorption barrier for radionuclides released
 Chemical barrier - once the capsule has been breached, can act to regulate the
availability of radionuclides for release into water through its impact on the solubility of
radionuclides
 Physical barrier – until the casing starts to degrade will limit the flow of water up into
borehole due to low permeability
 Until degraded, restricts the flow of water into the disposal zone in saturated systems
35
Near-field
Component
Disturbed zone
backfill
Closure zone
backfill
Post-closure Safety Related Functions
 Physical barrier – limits the flow of water into the borehole due to low permeability
 Physical barrier – once the disposal container and capsule have been breached, can act
as low permeability barrier to the migration of radionuclides from the borehole in liquid
and gaseous phases
 Chemical barrier - once the disposal container and capsule have been breached, can act
as sorption barrier for radionuclides released from the borehole
 Chemical barrier - once the capsule has been breached, can act to regulate the
availability of radionuclides for release into water through its impact on the solubility of
radionuclides
 Cement can passivate corrosion of stainless steel capsule and reduce chloride levels in
water through formation of calcium chloride
 Physical barrier – limits the flow of water into the borehole due to low permeability
 Physical barrier - inhibits disruption of the disused source by surface erosion, human
intrusion, and biotic intrusion
 Physical barrier – once the disposal container and capsule have been breached, can act
as low permeability barrier to the migration of radionuclides from the borehole in liquid
and gaseous phases
 Chemical barrier - once the disposal container and capsule have been breached, can act
as sorption barrier for radionuclides released from the borehole
 Chemical barrier - once the capsule has been breached, can act to regulate the
availability of radionuclides for release into water through its impact on the solubility of
radionuclides
 Cement can help maintain high pH conditions which then passivate corrosion of
stainless steel capsule and reduce chloride levels in water through formation of calcium
chloride
36
3.1.5 Uncertainties
The key uncertainty associated with the near field is the physical and chemical
characteristics of the sources and their ages and hence their current activity levels.
These uncertainties are now being addressed by NRWMC of GAEC through further
source characterisation work.
3.2 GEOSPHERE
There is currently an absence of site-specific geosphere data. However, two site
characterisation boreholes will be drilled in late 2011/early 2012 to allow sitespecific data to be obtained. In the absence of site-specific data, this first iteration of
the PCSA will use data on the regional geology and extrapolate to the site.
Subsequent iterations will be able to use the site-specific geosphere data collected
from the site characterisation programme.
3.2.1
Structural Geology and Stratigraphy
The site for the BDF is within the Accra Plains and overlaps the boundary between
the Togo Series and Dahomeyan System (Figure 7), both of which are of
Precambrian age. The Dahomeyan is the major bedrock formation underlying the
site. The Togo Series are predominantly composed of quartzite and phyllite while the
Dahomeyan System consists of quartzite, gneiss and schists. The Togo-Dahomeyan
boundary at this location is an ancient overthrust (actually, a group of parallel
thrusts) which also results in the two formations having an interleaved relationship.
Both formations are also intensely folded and have undergone various degrees of
metamorphism. Because of the overthrust and interleaving character, the younger
Togo series sometimes lies above and sometimes lies within the older Dahomeyan.
This tectonic deformation was caused by the Pan-african event which ended during
the Cambrian, around 500 million years ago. The overthrust is dipping at a shallow
angle (5-10 from horizontal) and has a surface signature in the order of 100 m in
width (IAEA, 2006).
The Dahomeyan system occurs essentially as alternating belts of acid and basic
gneisses. The acid Dahomeyan group consists of two alternating belts. The first belt
lies to the immediate east of the Togo-Akwapim ranges extending from the coastal
plains of Accra to Kpong in north-northeast direction. It encloses a series of
disconnected linear Togo quartzite outliers. The second acid gneiss belt is located
east of the metabasic rocks and stretches in a similar north-northeast direction from
the east of Prampram Rocks of the acid Dahomeyan consist generally of muscovitebiotite gneiss, quartz-feldspargneiss, augen gneiss and minor amphibolites. These
rocks decompose to slightly permeable calcareous clay. The basic Dahomeyan rocks
could also be subdivided into two groups, namely: the metabasics and the basic
intrusive (Darko et. al., 1995).
37
Figure 7: Geological map of the region immediately around the proposed site
38
20 km
GAEC site
Akwapim FZ
Superficials
Potential
host rock
Dahomeyan?
Birriminian?
Togo and
Dahomeyan
interleaved
Dahomeyan
Figure 8 Geology of Accra Plains showing Akwapim Fault zone and the
Proposed Site for the BDF
3.2.2 Seismicity
There are no records of any severe earthquakes. The Akwapim hills and the Togo
range mark the line of a major active fault zone that runs north-east to Lake Volta
and south-west to a major offshore east-west fault zone in the Gulf of Guinea: the
Coastal Boundary fault (Figure 9). Three seismic active zones in the region include,
two zones on the Coastal Boundary Fault (40 km south of Accra) and the area north
of Akosombo on the Akwapim range (85 km north-west of Accra). Though the
GAEC reactor has been designed to resist an earthquake of 0.23g of grade 8
intensity, this is not expected in the area.
The whole of the site is covered by loose unconsolidated and weathered material that
is generally a few metres deep but which sometimes extends to a considerable depth,
especially in the western part of the site. This, it has been suggested, may reflect the
presence of troughs formed by downfaulted blocks. This indicates the existence of
seismic activity in the geologically past and it probably results from movements
along the Akwapim fault line. The unfaulted Dahomeyan appears to be a very
competent rock that would not yield to the energy imposed by a movement of the
Akwapim range (IAEA, 2006). Though there is evidence of seismicity in the
geological past but there is no significant current day seismic activity.
39
Figure 9.StructuralGeology of the Accra Region
3.2.3
Hydrogeology
The water table at the site generally has a depth of between 3 and 15 metres. The
unconsolidated deposits close to the surface appear to form a high transmissivity
“active zone” that acts as the main conduit for moving groundwater. The presence of
a clay layer below the unconsolidated rocks appears to act as an aquitard, isolating
the active zone from the deeper rocks. Deeper groundwater appears to have lower
solute levels than the surface water, which again suggests a degree of hydraulic
isolation
The Togo and Dahomeyan rocks do not contain aquifers in the strict sense of the
word but may hold some water in joints and fissures particularly the Togo quartzites
that are often well jointed (Akaho et al., 2003).
Since the main rocks have very low permeability, groundwater occurrence in the
Accra plains is controlled mainly by the development of secondary porosities, e.g.
40
fractures, faults joints etc. and the associated weathered zone. Two types of aquifers
occur, i.e. the weathered zone aquifers and the fractured zone aquifers. The
weathered zone aquifers are either semi-confined or phreatic. The fractured zone
aquifers generally are mainly semi-confined or confined. Aquifer yields are also
highly variable (0.7-27.5 m3 hr-1) with a mean value of 2.7 m3 hr-1. Transmitivity
values are generally low due to the clayey content of the regolith. They vary from
0.23m2 hr-1 in the clayey regolith to 4.0m2 hr-1 in fissured zones (Kortatsi and
Jorgensen, 2001; WRRI, 1996).
3.2.4
Geochemistry
The data presented in Table 8 and Table 9 is the hydrochemical results from three
boreholes on the GAEC site and boreholes in the Accra plains, respectively.
Analysis of these data indicates that the water is brackish and the presence of NO3
suggests that we might have aerobic conditions. These however, would have to be
from measurements of Eh from the site investigation boreholes
41
Table 8: Chemical analysis of groundwater samples from boreholes on the GAEC Site
BH ID
BF1
BF2
BF3
Location
BNARI Farms
(GAEC)
BNARI Farms
(GAEC)
BNARI Farms
(GAEC)
Temp
oC
pH
Depth
TDS
Ele.
Con.
Ca
Mg
Na
K
HCO3
Cl
SO4
NO3
SiO2
Total
cations
Total
Anions
(m)
(mg/l)
(µS/cm)
(mg/l)
(mg/l)
(mg/l)
(mg/l)
(mg/l)
(mg/l)
(mg/l)
(mg/l)
(mg/l)
(meq/l)
(meq/l)
26.8
7.7
74
1671
3760
144.3
85.9
395
70
395.01
746.08
130.3
1.5
35.1
32.9195
30.25009
26
7.5
70
438
684
32.4
17.5
87.9
10.8
92.66
149.95
41.1
2.05
11.4
7.0831
6.638362
26.6
7.7
50
379
592
31.5
17.8
74.5
12.5
83.15
139.98
39.1
1.5
11.5
6.5316
6.150867
42
Table 9: Chemical Data of Representative Samples (from WRI Data Bank) Kortatsi and Jorgensen (2001)
43
3.2.5 Natural Resources
The site has no natural resources such as gold that require excavation by extensive
surface excavation or underground mining. There are also no significant sources of
geothermal heat or gas and oil in the vicinity.
There is however the possibility of the abstraction of groundwater. Groundwater is
abstracted from all geological formations in Ghana. The estimated annual
abstraction of groundwater based on 12h of pumping per day for the Accra plains is
2.5E+6 m3a-1 (Kortatsi, 1994).
3.2.6 Safety Related Functions
The geosphere has a number of safety-related functions; these are summarised in
Table 10.
TABLE10: GEOSPHERE COMPONENTS AND THEIR SAFETY RELATED
FUNCTIONS
System Component
Saturated Zone
Post-closure Safety Related Functions
 Physical barrier– lowers radionuclide concentrations in groundwater due
to dispersion and diffusion.
 Physical barrier– the depth of the disposal zone (50 m below the ground
surface) isolates the waste from intrusion (human and animal) and
geomorphological processes such as surface erosion.
 Chemical barrier– can retard the migration of radionuclides due to
sorption of radionuclides. This will cause greater radionuclide decay in
the saturated zone and so lower concentrations reaching the biosphere.
 Tectonic, seismic and geomechanical stability –helps maintain the
integrity of the BDF.
 Absence of economically viable mineral resources - limits the nature
and likelihood of human intrusion into the BDF
3.2.7 Uncertainties
There are uncertainties with regards to the precise nature of the stratigraphic and
hydrogeological conditions at the site as the two site characterisation boreholes
are yet to be drilled. In the absence of site-specific data, there are regional data
that can be used.
44
3.3 BIOSPHERE
3.3.1
Topography
The site of the BDF lies on a small hill on the south-eastern flank of the Akwapim
hills (Figure 3), part of the Togo range that extends north-eastwards for hundreds of
kilometres. The proposed disposal site location lies in the eastern part of the GAEC
site in an area that has been generally allocated to radioactive waste management.
This part of the site slopes gently down eastwards to the River Onyasia, a fall of less
than 10 metres but still above any known flood level.
A surface erosion rate for the Accra region of 1E-3 m y-1 can be calculated from data
provided in Oduro-Afriyie (1996), which could result in the disposal zone being
uncovered by erosion after 50,000 years (for a closure zone depth of 56.5 m – see
Section 2.2.3).
3.3.2
Climate
The climate of the plains is equatorial with two rainy and two dry seasons (Figure
10). There is a dry season from November to March during which rainfall is around
30 mm per month. This season is followed by a rainy season from April to June
during which an average of about 130 mm rain falls per month there is a little dry
season from July to August after which there is another rainy season. The mean
annual rainfall is 800 mm and the mean annual temperature is 26.5°C (Figure 11).
The daily variations of temperature reach between 5 °C and 6 °C. The mean annual
pan evaporation is in the order of 1800 mm.
Figure 10 Mean Monthly Rainfall
45
Figure 11: Mean Monthly Temperatures
Source: Ghana Meteorological Services Dept. Accra/SNC LAVALIN, 1995
3.3.3
Surface Water Bodies
The only major river near the BDF site is the river Onyasia. It is located 1.3 km from
the proposed BDF and drains southwards through Achimota village to Accra with a
measured flow velocity of 0.8 m s-1 (2.5E+7 m y-1), depth of 0.6 m and width of
6.8 m (site measurements from October 2011).The broad valley of the Onyasia river
flanks the site on its eastern margin, swampy conditions are generally found in the
north-east of the site. During the wet season, small localised swamps develop which
may persist well into the dry season. Surface run-off in this area is very low as the
top-soil is everywhere sandy. However, after heavy storms there may be some
movement of water over the clay horizon below the sandy top-soil (Akaho et al.,
2003).
The Atlantic Ocean is 30 km away from the site (Akaho et al., 2003) therefore there
is no adverse effect that is envisaged for the safety of the BDF.
3.3.4
Human Activity and Biota
Members of the general public are not allowed on the site as it is within GAEC’s
restricted site.
Outside the boundaries, the land is used for mainly small scale agriculture. The crops
grown are vegetables, maize, sugar cane and plantain. Animals reared include fowl,
goats, sheep, pigs and rabbits. There is a cattle ranch near the Kwabenya village
46
where cows are reared on a commercial scale. There are two vegetable plantations in
the area for growing vegetables on a commercial scale (Akaho et al., 2003).
Abstracted water from boreholes and the nearby Onyasia riveris used for domestic
(drinking) and agricultural purposes (watering of cattle and irrigation of vegetables).
In the Accra Plains, about 70% of the boreholes are drilled for agricultural purposes
and 33% of those are used for irrigation. Irrigation is limited to watering moderate to
high salt tolerant vegetables such as cabbage, onion, tomatoes and carrots (Kortatsi,
1994, Kankam-Yeboah, 1987). In the Accra Plains of Southern Ghana, a pilot project
for carrying out dry season vegetable farming with borehole water is currently being
carried out. Crop yields of 5 t ha-1 and 3 t ha-1 in the cases of cabbage and onions
have been realized (Kortatsi, 1994).
The major industries located at 15 km away from the site are a brewery and a
pharmaceutical manufacturing company. Stone quarries and other small scale
welding industries are located about 6 km away from the site (Akaho et al., 2003).
The activities of these quarrying operations are not likely to have any effect on the
BDF.
3.3.5
Near-surface Lithostratigraphy
The top-soil is everywhere sandy (Akaho et al., 2003). The soils are mainly
Vertisols. The vertisols of the Accra Plains, generally referred to as Tropical Black
Earth, is classified as Calcic Vertisol (Abunyewa et. al., 2004, FAO/UNESCO,1990).
They are agriculturally under-utilized within the traditional farming practices
because of constraints to crop production such as inefficient water and nutrient
management practices. Available nitrogen has been found to be generally low in the
Vertisols of the Accra Plains (Acquaye and Owusu-Bennoah, 1989).Vertisols are
characterized by very low basic water infiltration rate or low saturated hydraulic
conductivity because of their high smectite content (Dudal and Bramao 1965;
Coulombe et al., 1996) and, therefore, are susceptible to water logging during the
peak of the major rainy season.
3.3.6 Safety Related Functions
The biosphere is seen as pathways that can lead to exposure or impacts and no safety
related functions are assigned to it.
3.3.7 Uncertainties
The biosphere description given above is for the present-day biosphere. Over the
timescales of interest to the PCSA, the biosphere will be subject to change due to
environmental processes such as erosion, long-term climate change and human
processes (e.g. human activities – encroachment of housing over the last 10-20
years). These uncertainties are addressed in subsequent sections of this report.
47
4
IDENTIFICATION AND DESCRIPTION OF SCENARIOS
A scenario is a hypothetical sequence of processes and events, and is one of a set
devised for the purpose of illustrating the range of future behaviours and states of a
disposal system, for the purposes of evaluating a safety case (IAEA, 2004). Scenarios
handle future uncertainties associated with the processes and events by describing
alternative future evolutions of the disposal system and allow for a mixture of
quantitative analysis and qualitative judgements. The purpose of scenario
identification is not to try and predict the future; rather, it is to use scientifically
informed expert judgement to guide the development of descriptions of possible
future evolution of the disposal system to assist in making safety related decisions.
4.1 APPROACH
The approach to scenario identification and description used for this initial iteration
of the Ghana-specific PCSA is to take the scenarios identified in the IAEA Generic
Safety Assessment for the borehole disposal concept (IAEA 2008) and review and, if
necessary, modify them to produce site-specific scenarios. Using information relating
to the assessment context (Section 2), the system description (Section 3) and the
status of scenario-generating external factors8, the ‘Design Scenario’ identified in
IAEA (2008) was reviewed and modified. The scenario represents how the disposal
system can be expected to evolve assuming the borehole’s design functions as
planned and it provides a benchmark against which alternative scenarios can be
compared.
The four alternative scenarios that were then identified in IAEA (2008) by
considering possible alternative conditions for the scenario-generating external
factors (Table 11), were then reviewed taking account of the assessment context and
the system description provided in Sections 2 and 3. It was considered that the same
four alternative scenarios were applicable to the PCSA.
‘The Defect Scenario’ – it is assumed that not all components of the near field
perform as envisaged in the Design Scenario due to either defective
manufacturing of waste packages (e.g. welding defects), or defective
implementation in the borehole (e.g. improper emplacement of cement grout).
This results in the earlier release of radionuclides from the near field.
8Sub-divided
into repository factors, geological processes and events, climate processes and events, and future
human actions and behaviours.
48
TABLE 11. STATUS OF EXTERNAL FACTORS FOR ALTERNATIVE
SCENARIOS
External
Factors
Defect
Repository
Factors
Not all near-field
components perform
as envisaged in the
Design Scenario
Geological
Processes and
Events
No unexpected
features, processes or
events
Climate
Processes and
Events
Constant climate
conditions with
continuous, gradual
surface erosion
Future Human
Actions and
Behaviours
Domestic and
agricultural use of
water from an
abstraction borehole
sunk at the end of the
institutional control
period.
Construction of a
dwelling above the
disposal borehole at
the end of the
institutional control
period
Alternative Scenarios
Unexpected
Changing
Geological
Environmental
Characteristics
Conditions
Borehole
Borehole
constructed, operated constructed, operated
and closed as
and closed as
designed and planned designed and planned
Borehole
Disturbance
Borehole
constructed,
operated and
closed as
designed and
planned
No unexpected
Unexpected features, No unexpected
features, processes or features,
processes or events
events
processes or
events
Constant climate
Constant climate
Changing climate
conditions with
conditions with more conditions with
continuous, gradual
continuous,
rapid surface
surface erosion
gradual surface
erosion
erosion
Domestic and
Domestic and
Disturbance of
agricultural use of
agricultural use of
the disposal
water from an
water from an
borehole by
abstraction borehole
abstraction borehole
human intrusion
sunk at the end of the sunk at the end of the at the end of the
institutional control
institutional control
institutional
period.
period.
control period.
Construction of a
Construction of a
Domestic and
dwelling above the
dwelling above the
agricultural use
disposal borehole at
disposal borehole at
of water from an
the end of the
the end of the
abstraction
institutional control
institutional control
borehole sunk
period
period
immediately
adjacent to the
disturbed
disposal
borehole.
Construction of a
dwelling above
the disturbed
disposal
borehole.
Note: External factors in italic bold differ from those assumed for the Design
Scenario.
‘The Unexpected Geological Characteristics Scenario’ – it is assumed that the
actual performance of the geosphere from a safety perspective is worse than the
expected performance. The geosphere is subjected to an unexpected seismic
49
event resulting in the reactivation of high permeability fractures and
modification of associated sorption properties.
‘The Changing Environmental Conditions Scenario’ – it is assumed that the
disposal system is affected by climate change resulting in modifications to
certain geosphere characteristics (e.g. groundwater recharge rates) and
biosphere characteristics (e.g. water demand, surface erosion rates).

‘The Borehole Disturbance Scenario’ – it is assumed that drilling of a water
abstraction borehole immediately adjacent to the disposal borehole results in
the disturbance of the disposal borehole and the earlier release of radionuclides
from the near field and subsequent exposure of humans to radionuclides (e.g.
due to the use of contaminated water from the abstraction borehole).
On the basis of information provided in the assessment context and system
description and in light of the assumed status of the EFEPs, each of the above basic
scenario descriptions is developed further in the following sub-sections.
4.2 DESIGN SCENARIO
4.2.1
Description
The first stage in the further development of the Design Scenario description is to
consider the temporal evolution of the disposal system (i.e. the near field, geosphere
and biosphere). Each component of the disposal system is considered in turn.
The near field has been sub-divided into a series of components based on the system
description (Table 5) and the temporal evolution of each component considered. The
temporal evolution of each near-field component is documented in Table 12 together
with the associated assumptions. For the geosphere component of the disposal
system, there is no evolution over the assessment period since the site is located in a
geologically stable area with no or extremely limited tectonic and seismic activity
(see Section 3.2).
50
TABLE 12. TEMPORAL EVOLUTION OF THE NEAR-FIELD COMPONENTS
FOR THE DESIGN SCENARIO
Near-field
Component
Source
container
Capsule
Containment
barrier
Disposal
container
Disposal zone
backfill
Disposal zone
plug
Casing
Disturbed
zone backfill
Closure zone
backfill
Temporal Evolution
In most cases the source containers will still be intact at the time of disposal, due to proper
quality control and quality assurance procedures. However, the Na-22 sources to be
disposed are not sealed (see Table 3), and the longevity of the other source containers
cannot be guaranteed. Consequently, it is assumed that the source containers will have
failed prior to disposal. It is assumed that the radionuclides in the source containers are
available for potential release only once the capsule that surrounds the source container is
breached.
A number of different types of corrosion can occur including general and localised (e.g.
pitting and crevice). Corrosion of capsule is assumed to start only once the disposal
container and the associated containment barrier has been breached by water (see below).
Physical and chemical degradation of the cement grout will start only once the disposal
container has been breached and the cement grout is contacted by water. Initially the
hydraulic conductivity might decrease due to carbonation, however with time it will
increase due to the physical (e.g. cracking) and chemical (e.g. calcium leaching and
sulphate attack) degradation of the cement grout due to contact with flowing water.
Chemical degradation generally results in a decrease in the cement grout’s sorption
capacity.
See discussion concerning capsule for corrosion mechanisms. Corrosion of disposal
container is assumed to start before the corrosion of the capsule.
It is assumed that any shrinkage or jointing cracks that might form in the cement grout
backfill do not act as significant water flow and hence radionuclide migration pathways.
Initially the hydraulic conductivity might decrease due to carbonation, however with time
it will increase due to the physical (e.g. cracking) and chemical (e.g. calcium leaching and
sulphate attack) degradation of the cement grout due to contact with flowing water,
especially once the borehole casing starts to fail (see below). Chemical degradation
generally results in a decrease in the sorption capacity of the cement grout.
Assumed to behave in the same manner as the disposal zone backfill.
Processes such as embrittlement, cracking and biodegradation are assumed to result in the
failure of the HDPE casing. Koerner et al. (2005) and Wienhold and Chudnovsky (2006)
suggest HDPE lifetimes in the region 100 to 400 years. However, there is considerable
uncertainty over lifetimes and it is therefore conservatively assumed that the casing fails
immediately following closure
Assumed to behave in the same manner as the disposal zone backfill.
The closure zone backfill will be subjected to surface erosion at a rate of about 1E-3 m y-1.
The characteristics of the native soil/crushed rock used to fill the first 5 m of the closure
zone from the ground surface is assumed to remain constant. The cement grout used to fill
the remainder of the closure zone is assumed to behave in the same manner as in the
disposal zone.
For the biosphere component of the disposal system, it is recognised that certain
changes might occur due to the effects of climate change. However, for the purposes
of this initial PCSA, it is assumed that such changes will not be significant.
Consistent with the recommendations of ICRP (2000) no consideration is given to
the development of new societal structures and technologies. Furthermore, consistent
with IAEA guidelines for the siting of radioactive waste disposal facilities (IAEA,
1994), the BDF is not located in an area
51
of significant geomorphological activity (including flooding), although a constant
rate of surface erosion is assumed (Section 3.3).
Consistent with the above discussion and the information in the assessment context
and system description, the following description of the Design Scenario can be
developed.
Construction, Operation and Closure Periods
The current assessment only assesses post-closure safety. This section is included to
clarify the status of the facility following construction, operation and closure. It is
assumed that the borehole is constructed, operated and closed as designed and
planned (see Section 3.1.2) with appropriate quality assurance and no accidents or
unplanned events.
During operations, measures are taken to ensure that the waste packages are
emplaced in a dry environment, and that shrinkage cracks in the backfill are
minimised. The whole site area is controlled to prevent animal and unauthorised
human access. All site investigation activities are managed with the intention to
ensure that there are no adverse effects on post-closure safety.
Institutional Control Period
Throughout the institutional control period of 50 years, a limited level of
environmental monitoring will be performed for the purpose of public assurance. All
monitoring activities will be managed with the intention to ensure that there are no
adverse effects on post-closure performance.
At closure, no markers, which might encourage deliberate human intrusion, are fixed
at the site to reveal the location of a radioactive waste disposal facility. However, a
detailed record of the disposal site as well as the disposal facility and its content will
be kept at the municipal assembly to enforce controls of the use of the land covering
the site. These land use controls are related to the erection of buildings at the site and
drilling of boreholes. After the institutional control period (50 years for the Design
Scenario – see Table 2), all societal memory of the site is assumed to be lost.
Following construction, it is assumed that groundwater starts to enter the borehole
and some corrosion of the stainless steel disposal containers begins. Nevertheless, the
containers remain intact and ensure that water does not come into contact with the
waste and there are no releases of gases.
Post-Institutional Control Period
Due to the corrosion of the stainless steel disposal containers and the subsequent
corrosion of the capsules, water eventually contacts the waste in source container,
which is assumed to have failed prior to disposal. The radionuclides in the source
container could be in a number of different physical and chemical forms (yet to be
determined) and release of radionuclides could occur in the liquid or gas phase.
52
For radionuclides released in the liquid phase, transport from the source container
through the various components of the near field can occur by advection, dispersion
and diffusion. The relative importance of these processes depends upon the
hydrogeological conditions at the site. Migration through the near field is limited by
decay/in-growth, and sorption of the radionuclides onto the cement grout in the near
field. On leaving the near field, the radionuclides migrate through the geosphere by
advection, dispersion and diffusion and are subject to decay/in-growth, and
retardation due to sorption onto the rocks. Flow can be through pores or fractures and
diffusion can occur into stagnant water in the rock matrix. Again the relative
importance of these geosphere processes depends on the hydrogeological conditions
at the site. The groundwater is assumed to be abstracted from the geosphere via an
abstraction borehole. The borehole is assumed to be down the hydraulic gradient
from the BDF and used for domestic purposes (drinking) and agricultural purposes
(watering of animals and irrigation of crops) (Section 3.3). The water is not treated or
stored before use. The main features of the Design Scenario for radionuclides
released in the liquid phase into the saturated disposal zones are summarised in
Figure 12.
Disposal
Borehole
Closure
Zone
56.5m
Water
Abstraction
Borehole
Watertable
Contaminant Plume
Waste
Disposal
Zone
43.5m
Groundwater Flow
Direction
Figure 12.Design Scenario: Liquid Releases
Contaminated groundwater might also discharge to the Onyasia river which could
also be used for domestic purposes (drinking) and agricultural purposes (watering of
animals and irrigation of crops). However, for the purposes of this initial PCSA,
discharge via the water abstraction borehole is considered since radionuclide
concentrations will be higher in the well water due to it closer proximity to the BDF
resulting in less dilution, dispersion, decay and absorption in the geosphere.
53
The failure of the containers and capsules could allow any radioactive gases to be
released. None of the sources currently contain radioactive gas (see Table 4),
however Rn-222 will ingrow from Ra-226. The very short half-life of Rn-222
(around 3 days) means that there is likely to be significant decay within the saturated
zone, and very little will reach the unsaturated zone and even less eventually
discharge into the biosphere. Furthermore, the Rn-222 could be dissolved into
groundwater rather than remain in gaseous form. So for the purposes of the current
assessment, no gaseous releases are considered.
The combination of the surface erosion rate (1E-3 m y-1 – see Section 3.3) and the
depth of the disposal zone from the ground surface (56.5 m – see Section 3.1.2)
results in the waste being uncovered after 50,000 years. The main features of the
Design Scenario for radionuclides released in the solid phase are summarised in
Figure 13.
Original Ground Surface
Closure
Zone
Eroded
56.5m
Soil Contaminated by Material
from Eroded Borehole
Waste
Disposal
Zone
Disposal Borehole
43.5m
Figure 13: Design Scenario: Solid Releases
4.2.2
FEP Screening
The Generic Safety Asssessment of the borehole disposal concept includes a
screening of FEPs on the basis of information provided in the assessment context;
system description and the scenario description (see Appendix D and E of IAEA
(2008)). A similar screening should be undertaken for the PCSA. Time constraints
associated with the development of the current iteration of the PCSA have meant that
54
this has not yet been undertaken. However, it is planned that future versions of the
PCSA will include such as screening.
4.3 DEFECT SCENARIO
4.3.1
Description
The scenario assumes that a properly qualified team applies appropriate quality
assurance and quality control (QA/QC) to the construction, operation and closure
activities. For example radiographic and other post-weld inspection procedures are
expected to be part of the waste capsule and container fabrication process. This
assumption of appropriate QA/QC limits the extent of the defects that might arise.
However, as in any engineering system, some defects may arise despite best efforts
to eliminate them. Furthermore, maintaining quality during field welding, as is
envisaged for the borehole disposal concept, is generally more challenging than
during shop welding.
Therefore the scenario assumes that not all components of the near field perform as
envisaged in the Design Scenario, resulting in the earlier release of radionuclides
from the near field.
A range of possible defects involving one or more of the near-field barriers (i.e.
capsule, containment barrier, disposal container, disposal and disturbed zone backfill
and casing) can be identified. These are summarised and screened in Table 13. Four
Defect Scenario variants are identified:
D1: all welds are satisfactory due to QA/QC except for the closure weld in one 316 L
waste container. All other near-field barriers as per Design Scenario.
D2: All welds are satisfactory due to QA/QC except for the closure weld in one 304
waste capsule. All other near-field barriers as per Design Scenario.
D3: degraded/incomplete disposal/disturbed zone cement grout. All other nearfield barriers as per Design Scenario.
D4: all welds are satisfactory due to QA/QC except for the closure weld in one 316 L
waste container and one 304 waste capsule. The faulty capsule is in the
faulty container. All other near-field barriers as per Design Scenario.
4.3.2
FEP Screening
As is the case with the Design Scenario, time constraints have meant that FEP
screening for the Defect Scenario has not yet been undertaken. However, it is
planned that future versions of the PCSA will include such screening.
55
TABLE 11. POSSIBLE DEFECTS CONSIDERED IN THE DEFECT SCENARIO
Description
Considered in
Defect Scenario
Calculations
(Variant
Number)
All welds okay due to QA/QC
except for the closure weld in
one 316 L disposal container.
Yes (D1)
Yes (D2)
Feasible although considered to be lower
consequence than D1. A similar probability for
an individual defect (i.e., 10-3) is assumed, and
for the probability that only 1 out of the 50 waste
capsules will contain a defect.
Yes (D3)
Cannot be ruled out. Probability assumed to be c.
1%.
All other near-field barriers as
per Design Scenario.
Missing/degraded/incomplete
disposal/disturbed zone cement
grout.
Cannot be ruled out.
For the mass production of welded structures
under strict QA/QC procedures, the probability of
an individual undetected, through-wall defect is
of the order of 10-3-10 -4(Doubt 1984)Because of
the potential for more challenging conditions for
welding and inspecting the waste container, the
higher end of this range (10-3) is assumed for the
PCSA. The probability that the weld on 1 of the
50 disposal containers in the borehole contains a
defect can be estimated based on a binomial
distribution, and is found to be 0.05 for an
individual probability of 10-3. The probability
that 2 out of the 50 disposal containers in a
borehole will contain defects is 0.0012 and is
considered to be too small to be of concern here.
All other near-field barriers as
per Design Scenario.
All welds okay due to QA/QC
except for the closure weld in
one 304 waste capsule.
Justification
Consider more rapid chemical and physical
degradation of cement grout than for Design
Scenario.
All other near-field barriers as
per Design Scenario.
Case of missing cement grout is covered under
“What-if” calculation presented separately from
the Defect Scenario results.
Missing casing.
No
No credit is taken for the casing in the Design
Scenario
Yes (D4)
Based on probabilities of 10-3 for individual weld
defects and of 0.05 that 1 out of the 50 disposal
containers and waste capsules contain a defect,
the probability that the defected waste capsule is
All other near-field barriers as
per Design Scenario.
All welds okay due to QA/QC
except for the closure weld in
one 316 L disposal container
and one 304 waste capsule. The
56
Description
Considered in
Defect Scenario
Calculations
(Variant
Number)
faulty capsule is in the faulty
container.
All other near-field barriers as
per Design Scenario.
Justification
inside the defected disposal container is 0.05 x
0.05  50 = 5 x 10-5. Although this is of low
probability, the consequences of this scenario
could be high (since there could be immediate
release from the waste package) and warrants
analysis.
This is considered to be the most likely twobarrier failure scenario.
4.4 UNEXPECTED GEOLOGICAL CHARACTERISTICS SCENARIO
This scenario assumes that the actual performance of the geosphere from a safety
perspective is worse than its expected performance, resulting in the more rapid
transport of radionuclides through the geosphere. This could be due to a number of
factors such as: higher hydraulic conductivities than anticipated; lower geosphere
sorption coefficients than anticipated; the presence of undetected high permeability
zone(s); and the reactivation of high permeability zone(s) due, for example, to
unexpected seismic activity.
It is not necessary to develop a separate scenario, as the additional geosphere
parameter sensitivity analysis (presented in Section 6.2) bound the consequences of
this scenario.
4.4.1
Changing Environmental Conditions scenario
This scenario assumes that the disposal system is affected by climate changes.These
changes could result in modifications to certain geosphere characteristics (e.g.
groundwater recharge rates) and biosphere characteristics (e.g. water demand,
surface erosion rates).
It is not necessary to develop a separate scenario, as the geosphere and biosphere
parameter sensitivity analysis (presented in Section 6.2) bound the consequences of
this scenario. Furthermore, results from a previous GSA that did consider an
environmental change scenario (Little et al., 2004), further support the screening out
of this scenario.
57
4.4.2
Borehole Disturbance Scenario
The impact of deliberate human intrusion is considered to be beyond the scope of the
PCSA (see Section 1.3).
The depth of the disposal zone (56.5 m thick from the ground surface), the small
footprint of the disposal borehole, and its location in an area that has no natural
resources requiring excavation by extensive surface excavation or underground
mining (Section 3.2), all mean that the likelihood of inadvertent human intrusion
directly affecting the disposal borehole is extremely low. Even if the site were to be
developed, given the disposal borehole’s narrow cross-sectional area (about 5E-2 m2)
and a site investigation borehole density of 1 per 1000 m2 (BSI, 1999), the likelihood
of an investigation borehole being within the footprint of the disposal borehole is
around 1 in 20,000. Furthermore, even if the investigation borehole were to be within
the footprint of the disposal borehole, the various components of the near field, such
as the steel of the disposal container and the capsule and the anti-intrusion barrier
above the disposal zone (see Section 3.1.2), could be expected to deter direct
intrusion into the disposal zone. Due to these reasons, further consideration is not
given to the borehole disturbance scenario in the current PCSA.
58
5
DEVELOPMENT AND IMPLEMENTATION OF MODELS
5.3 APPROACH
The model development and implementation process is shown in Figure 14.
Information from the assessment context, system description and scenario
development steps of the safety assessment approach can be used to help generate
conceptual models of the disposal system for the scenarios to be assessed (i.e. the
Design and Defect Scenarios). These conceptual models and their associated
processes are represented in mathematical models that are then implemented in
computer codes. Throughout this process, data are used to help develop the
conceptual and mathematical models and as input to the computer codes.
Assessment
Context
System Data
System
Description
Scenarios
Conceptual
Models
Model
Formulation
and
Understanding
Mathematical
Models
Implementation
Implementation of
(Step 4 of the
ISAM Safety
Assessment
Approach)
Model
Parameter
Mathematical
Values
and
Models in
Computer Tool(S)
Figure 14. Model Formulation and Implementation Process Used
59
5.2
CONCEPTUAL MODELS
The Interaction Matrix approach has been used to help identify the main components
of the disposal system and the processes that result in the release and migration of
radionuclides through the system (Appendix B). The conceptual model for each of
the system’s main components (near field, geosphere and biosphere) is summarised
below.
5.2.1 Near Field
The near field is comprised of a series of engineering barriers. Working from the
outside inwards, these comprise (see Figure 5):

the disturbed zone cement grout backfill;

the HDPE casing;

the disposal zone cement grout backfill;

the stainless steel disposal container;

the cement grout containment barrier inside the disposal container;

the stainless steel capsule; and

the source container.
The HDPE casing and source container are assumed to have failed by closure of the
borehole (Table 12). Therefore, the migration of radionuclides from the near field is
controlled by the degradation of the cement grout and stainless steel barriers and the
release of radionuclides from the disused source into the borehole. The models for
cement grout and stainless steel degradation adopted for the PCSA are consistent
with those described in detail in Appendices H and I of IAEA (2008), respectively,
and summarised below. The release and transport models are also presented below.
Cement Grout Degradation
The various alteration processes discussed in Appendices E and H of IAEA (2008)
(e.g. chloride binding, carbonation, ettringite precipitation, expansion caused by
corrosion) will affect the chemical and physical degradation of the cement grout.
Consistent with IAEA (2008), four stages of degradation are considered based on the
work reported in Berner (1992) and Berner (2004).

Stage 1 - porewater pH is around 13.5, owing to the presence of significant
NaOH and KOH and such high pHs can persist during flushing by about 100
pore volumes of water. It is assumed that the values for chemical and physical
parameters such as sorption coefficient, porosity and hydraulic conductivity are
comparable with those for undegraded cement grout.
60

Stage 2 - porewater pH has fallen slightly to about 12.5, owing to buffering by
Ca(OH)2 and this pH can persist during flushing by an additional 900 pore
volumes. Although pH has declined slightly, it is assumed that the chemical
and physical parameter values are the same as for Stage1.

Stage 3 - porewater pH diminishes steadily from 12.5 to about background
groundwater pH, owing to buffering with C-S-H phases having progressively
decreasing Ca/Si ratios. This stage can persist during flushing by approximately
an additional 4000 to 9000 pore volumes. There is significant chemical and
physical degradation of the cement grout resulting in changes in chemical and
physical parameter values. It is assumed that there is a linear change during
Stage 3 in parameter values from the start value (i.e. value for undegraded
conditions) to the end value (i.e. value for degraded conditions).

Stage 4 - porewater pH returns to that of the background waters and the cement
grout is fully degraded. The chemical and physical parameter values are the
same as those at the end of Stage 3 (i.e. degraded values).
As discussed in Appendix H of IAEA (2008), the exact duration of each stage
depends on the composition of groundwater (in particular groundwater pH), the rate
of groundwater flow (the higher the flow, the more rapid the pore flushes and the
more rapid the degradation) and the nature of the scenario assessed. Shorter stages
are assumed for the Defect Scenario Variant D3 (incomplete or degraded disposal
zone cement grout) (see Section 4.3).
Stainless Steel Corrosion
Given the potential aerobic groundwater at the site, stainless steel could to be subject
to general corrosion, as well as localised corrosion, in the form of crevice corrosion
or pitting, and stress corrosion cracking under certain conditions (Figure 15).
Microbiologically influenced corrosion (MIC) of stainless steel is also possible in
natural groundwaters, but because of the conditioning of the near-field pH by the
cementitious materials, microbial activity will be limited until such time that the
near-field pH drops below ~pH 10. Since the majority of containers are calculated to
have failed by general corrosion before the pH drops below this value, MIC has not
been explicitly included in the corrosion model developed for the PCSA.
61
Is the environment
aerobic or anaerobic?
Aerobic
Is the pH less than
the critical pH for localized
corrosion?
Anaerobic
General
corrosion
only
Yes
No
General
corrosion
only
No
General
corrosion
only
Is the Cl- concentration
greater than the critical
concentration for localized
corrosion?
Yes
General and
localized corrosion
and SCC
Figure 15: Decision Tree for the Corrosion Model Used for the Borehole
Disposal Concept Generic Safety Assessment.
For the corrosion model, a four-stage time-dependent evolution of the near-field
chemistry has been used consistent with that used in IAEA (2008). The evolution of
the cement grout porewater pH is assumed to evolve through the stages defined
above. The corrosion model assumes that the corrosion rate is a function of not only
pH but also chloride concentration and redox potential (reducing low chloride
conditions give lower corrosion rates than oxidising high chloride conditions). The
porewater chloride concentration and redox potential are assumed to be spatially and
temporally constant and are consistent with the groundwater concentrations given in
Section 3.2.4.
62
Table 14 provides a summary of the corrosion processes included in the model for
each stage in the evolution of the environment and for the site groundwater taking
into account that it is currently uncertain whether conditions are aerobic or anaerobic
at the site. It is assumed that internal corrosion of the disposal containers and
capsules is not significant and is therefore not considered.
The four Defect Scenario variants identified in Section 4.3 will reduce the lifetimes
of the affected containers due to the earlier onset of corrosion, although the processes
will be the same as the Design Scenario. Variant D3 (incomplete or degraded
disposal and disturbed zone cement grout) compromises the ability of the cement
grout to condition the near-field pH.
TABLE 12. SUMMARY OF THE PCSA CORROSION MODEL
pH
Aerobic conditions
Anaerobic conditions
Stage 1 (pH 13.5)
General corrosion only
General corrosion only
Stage 2 (pH 12.5)
General corrosion only
General corrosion only
Stage 3(a)
General corrosion only
General corrosion only
(pHCRIT< pH < 12.5)
Stage 3(b)
General and localised
General corrosion only
corrosion
(pHGW< pH  pHCRIT)
Stage 4
General and localised
General corrosion only
(pHGW)
corrosion
Note
pHCRIT is defined as pH 10 for Type 316 stainless steel and pH 11 for Type 304 (see
Appendix I of IAEA 2008).
Anaerobic corrosion is accompanied by the generation of hydrogen gas. The rates of
anaerobic corrosion are lower and are estimated to be in the range 0.01-1 m y-1, the
lower end of the range corresponding to fresh, high-pH conditions and the upper end
of the range to saline, near-neutral pH waters. Because the disposal containers tend
to fail prior to the establishment of near-neutral pH conditions, the predicted
maximum rate of H2 generation is of the order of 4-8 ml y-1 per disposal container, or
200-400 ml y-1 for the entire borehole. Following failure of the disposal containers,
the rate of gas production will decrease by a factor of ~12 (for the same corrosion
rate), as the surface area of the waste capsule is much smaller than that of the
disposal container. It is likely that H2 generated at these rates will be transported
away from the borehole and that a separate gaseous H2 phase is unlikely to develop
within the borehole.
Release of Radionuclides
Due to the corrosion of the stainless steel disposal containers and the subsequent
corrosion of the capsules, water eventually contacts the waste in source container,
which is assumed to have failed prior to disposal. The radionuclides in the source
container could be in a number of different physical and chemical forms and release
of radionuclides could occur on breaching of the waste capsule due to the following
mechanisms.
63

Instantaneous dissolution of radionuclides that are likely to be in a form that
would result in immediate release to water (e.g. soluble solid, surface
contamination) (i.e.Sr-90, I-129, Cs-137, Ra-226, Am-241and Cf-252).

Congruent release of radionuclides that are likely to be in a form that would
result in slow release to water (e.g. solid with low solubility) (i.e. Co-60 due
to its typically low solubility and metallic waste form – see Table 15 of IAEA
(2008)).
It is recognised that the instantaneous dissolution and congruent release mechanisms
could, under certain circumstances, be solubility limited. However, no solubility
limitation is considered for the reference case calculations (a conservative
assumption).
Migration of Radionuclides
For radionuclides released in the liquid phase, transport from the source container
through the various components of the near field can occur by advection, dispersion
and diffusion. The relative importance of these processes depends upon the
hydrogeological conditions at the site. Migration through the near field is limited by
decay/in-growth, and sorption of the radionuclides onto the cement grout in the near
field. It is assumed that the migration is not solubility limited.
For radionuclides released in the solid phase due to erosion of the closure zone, it is
assumed that the radionuclide in the topmost container is transferred directly into the
soil once the closure zone has been eroded (i.e. 50,000 years).
The associated near field migration processes are summarised in the yellow boxes in
Figure 16. Note that the processes considered for the Defect Scenario are the same
as those for the Design Scenario since the faster degradation rates, earlier failure
times, and faster radionuclide migration times of the Defect Scenario can be
accounted for by modifying the associated parameters in the mathematical model
(e.g. container degradation rates) rather than considering different processes. For
disposal in the saturated zone, the position of the defective capsule/container is not
important since the flow from the disposal borehole to the abstraction borehole is
assumed to be horizontal.
5.2.2
Geosphere
On leaving the near field, the radionuclides in groundwater migrate through the
geosphere by advection, dispersion and diffusion and are subject to decay/in-growth,
and retardation due to sorption onto the rocks. For BDF’s geosphere, flow occurs
through pores or fractures and diffusion occurs into the rock matrix. The
groundwater is assumed to be abstracted from the geosphere via an abstraction
borehole that is drilled 100 m down the hydraulic gradient from the disposal borehole
once institution controls are assumed to be no longer effective (i.e. 50 years after
64
closure). The associated geosphere migration processes are summarised in the blue
boxes in Figure 16.
5.2.3
Biosphere
The groundwater abstraction borehole is assumed to be used for domestic purposes
(drinking) and agricultural purposes (watering of cows and irrigation of root and
green vegetables) consistent with current practice in the vicinity of the site (Section
3.3). The water is not treated or stored before use. Humans are exposed via ingestion
of water, animal products and crops, inadvertent ingestion of soil, external irradiation
from soil, and inhalation of dust.
For radionuclides released in the solid phase due to erosion of the closure zone, it is
assumed that the contaminated soil is used for the growing of vegetables by a site
dweller. Humans are exposed via ingestion of vegetables, inadvertent ingestion of
soil, external irradiation from soil, and inhalation of dust.
The associated biosphere migration processes are summarised in the green boxes in
Figure 16.
65
Figure 16. Interaction Matrix for the Design Scenario
1
Source
(degradation,
dissolution, decay)
2
Advection
Dispersion
Diffusion
Groundwater flow
(only once capsule is
breached)
Containment
Barrier
(degradation,
sorption,
decay)
Groundwater flow
(only once
disposal container
is breached)
3
4
5
6
7
A
B
C
Disposal Zone
(degradation,
sorption,
decay)
Groundwater flow
Diffusion
Advection
Dispersion
Diffusion
Closure Zone
(degradation,
sorption,
decay)
Advection
Dispersion
Diffusion
10
11
12
13
Advection
Dispersion
Diffusion
Disturbed Zone and
Plug
(degradation,
sorption,
decay)
Unsaturated
Zone
F
Groundwater flow
Recharge
Saturated
Zone
(advection,
dispersion,
diffusion,sorp
tion, decay)
G
Percolation
Irrigation
Soil
(sorption,
decay)
Precipitation
Deposition
Death
and
decay
H
I
J
K
L
9
Advection
Dispersion
Diffusion
D
E
8
Transfer into
soil due to
erosion of
closure zone
Abstraction
Excretion
Bioturbation
Ploughing
Irrigation
Suspension
Root uptake
Atmosphere
Deposition
Ingestion
Crops
(translocation)
Cultivation
Harvesting
Ingestion
Advection
Dispersion
Diffusion
External
irradiation
Ingestion
Inhalation
Erosion
Percolation
Ingestion
Food
preparation
losses
Animals
Ingestion
Rearing
Humans
Dispersion
Excretion
Elsewhere
(decay)
M
66
5.3 MATHEMATICAL MODELS
Mathematical models translate the assumptions of a conceptual model into the formalism
of mathematics, usually sets of coupled algebraic, differential and/or integral equations
with appropriate initial and boundary conditions in a specified domain. These equations
are solved by computer software to give the temporal and spatial dependence of the
quantities of interest (such as radionuclide concentrations and doses to humans).
For the PCSA, an assessment model has been developed to allow the calculation of the end
points identified in Section 2.4. In addition, two supporting models have been developed
to represent the degradation of the cement grout and the corrosion of the containers in
detail and to provide associated input into the assessment model. The assessment and
supporting models are discussed in turn below.
5.3.1
Assessment Model
It was decided to implement the assessment model in the most recent version of the
AMBER software tool (version 5.4) (Quintessa, 2010) since it is a suitable tool in which to
implement the conceptual models developed in Section 5.2. Furthermore, it has been used
to develop models for the Generic Safety Assessment of the borehole disposal concept
(IAEA, 2008). AMBER uses a compartment model approach to represent the migration
and fate of contaminants in the disposal system. The use of AMBER places two main
conditions on the mathematical representation of a disposal system.
The first condition is that the system has to be discretised into a series of compartments.
Using the compartment modelling approach, a disposal system may be represented by
discretising it into compartments that can correspond to the components identified in the
conceptual model. It is assumed that either uniform mixing occurs over the timescales of
interest, or the distribution of the contaminant within the compartment is not important so
that a uniform concentration over the whole compartment can be used either for
subsequent transport or for deriving end points of interest. Therefore each compartment
should be chosen to represent a system component for which one or other of these
assumptions is reasonable.
The second condition is that processes resulting in the transfer of contaminants from one
compartment (the donor compartment) to another (the receptor compartment) need to be
expressed as transfer coefficients that represent the fraction of the activity in a particular
compartment transferred from the donor compartment to the receptor compartment per unit
time. The mathematical representation of the inter-compartmental transfer processes takes
the form of a matrix of transfer coefficients that allow the compartmental amounts to be
represented as a set of first order linear differential equations. For the ith compartment, the
rate at which the inventory of radionuclides in a compartment changes with time is given
by:
Equation 1

 

dN i
    ji N j   N M i  S i (t )     ij N i   N N i 
dt
 j i
  j i

67
wherei and j indicate compartments, N and M are the amounts (Bq) of radionuclides N and
M in a compartment (M is the precursor of N in a decay chain). S(t) is a time dependent
external source of radionuclide N (Bq y-1). Transfer and loss rates are represented by λ. λN
is the decay constant for radionuclide N (y-1) and λji and λij are transfer coefficients (y-1)
representing the gain and loss of radionuclide N from compartments i and j. For
simplicity, the above equation assumes a single parent and daughter. However, AMBER
allows the representation of multiple parents and daughters.
The solution of the matrix of equations given above provides the time-dependent inventory
of each compartment. Assumptions for compartment sizes then result in estimates of
concentrations in the corresponding media from which doses/intakes can be estimated.
The mathematical equations used to represent the release and migration processes and the
exposure mechanisms identified in the Interaction Matrices are described in Appendix C.
5.3.2
Supporting Models
As noted above, two supporting model have been developed to provide input data for use
in the assessment model.
The first has been developed in an Excel spreadsheet and has been used to calculate the
duration of each of the cement grout degration stages identified in Section 5.2.1 for the
hydrogeological and geochemical conditions considered in the PCSA. The model is
consistent with that described in Appendix H.4 of IAEA (2008). Using this model and the
geosphere characteristics described in Section 3.2, the cement degradation times given in
Tables 15 and 16 have been calculated.
Table 15 Cement Degradation Times for the Design Scenario and Defect Scenario D1,
D2 and D4
Geosphere
Duration (y)
Stage 1
Stage 2
Stage 3
Cumulative
Aerobic
1.03E+02
4.12E+02
4.91E+01
5.64E+02
Anaerobic
1.03E+02
9.27E+02
9.98E+01
1.13E+03
Aerobic
4.19E+01
1.67E+02
2.00E+01
2.29E+02
Containment
Barrier
Cement for
Small
Capsule (1)
Aerobic
4.19E+01
1.67E+02
2.00E+01
2.29E+02
Anaerobic
4.19E+01
3.77E+02
4.06E+01
4.59E+02
Containment
Aerobic
6.49E+00
2.60E+01
3.10E+00
3.56E+01
Backfill
Cement
68
Barrier
Cement for
Large
Capsule (1)
Anaerobic
6.49E+00
5.84E+01
6.29E+00
7.12E+01
Note
1.
Time from time of failure of the disposal container
Table 16 Cement Degradation Times for the Defect Scenario D3
Geosphere
Duration (y)
Stage 1
Stage 2
Stage 3
Cumulative
Backfill
Cement
Aerobic
5.15E+01
2.06E+02
2.46E+01
2.82E+02
Anaerobic
5.15E+01
4.64E+02
4.99E+01
5.65E+02
Containment
Barrier
Cement for
Small
Capsule (1)
Aerobic
2.09E+01
8.37E+01
9.98E+00
1.15E+02
Anaerobic
2.09E+01
1.88E+02
2.03E+01
2.30E+02
Containment
Barrier
Cement for
Large
Capsule (1)
Aerobic
3.25E+00
1.30E+01
1.55E+00
1.78E+01
Anaerobic
3.25E+00
2.92E+01
3.15E+00
3.56E+01
Note
1.
Time from time of failure of the disposal container
The second supporting model has also been developed in the same Excel spreadsheet and
used to calculate the failure times of the disposal container and the waste capsule for the
hydrogeological, geochemical and cement grout degradation conditions considered in the
PCSA. The model is consistent with that described in Appendix I of IAEA (2008). The
assumed general corrosion rates used are given in Table 17 based on data given for
Groundwater IDs 1 (aerobic) and 6 (anaerobic) in Table I.11 of IAEA (2008). These two
groundwater IDs are considered to be closest to the conditions at the BDF site. Consistent
with Appendix I.4.3 of IAEA (2008), it is assumed that the container will fail 100 years
after the initiation of localised corrosion. The failure times for the disposal container and
waste capsule for the Design Scenario are given in Table 18.
69
Table 17: Rates of General Corrosion Used in the Corrosion Model.*
Stage 3(a)
Stage 3(b)
(pHGW<
pH 
pHCRIT)
Stage 4
(pH 12.5)
(pHCRIT<
pH <
12.5)
0.1
0.1
0.1
0.5
1
0.02
0.02
0.02
0.05
1
Stage 1
Stage 2
(pH 13.5)
Aerobic
Anaerobic
Geosphere
(pHGW)
* Rates in m y-1
Table 18: Disposal Container and Waste Capsule Failure Times for the Design
Scenario
Geosphere
Disposal Container
Failure Time (y) (1)
Waste Capsule Failure Times (y) (1)
Small Capsule
Large Capsule
Aerobic
6.40E+2
9.55E+2
7.73E+2
Anaerobic
5.91E+3
8.75E+3
9.18E+3
Notes
1.
Time from time of waste emplacement.
5.4
DATA
The tables in Appendix D provide data for each of the parameters of the assessment model
described in Appendix C. Data relating to the inventory, borehole and its design and the
associated geosphere and biosphere characteristics have been drawn from the system
description (Section 3). Other radionuclide/element dependent and independent data have
been drawn from a number of relevant sources such as previous safety assessments (e.g.
Little et al., 2004) and data compilations (e.g. IAEA, 1994). Source references are given at
the end of each table in Appendix D.
5.5
IMPLEMENTATION
As mentioned in Section 5.3, the AMBER software tool was used to implement the
assessment model. The mathematical model and data described in Appendix C and D were
encoded directly into AMBER and quality assurance checks undertaken to ensure that the
implementation was correctly performed. The time dependent solution method used by
AMBER is described in Byrne and Hindmarsh (1975) and Robinson (2001). The
verification of the solution is discussed in (Quintessa, 2010)
In implementing the models and data in AMBER, the aim was to minimise the number of
input files that needed to be created and thereby reduce input error, facilitate checking and
70
updating, and avoid the replication of data needed by all or most calculation cases (e.g.
decay rates and dose coefficients). This was achieved through the use of a series of
“literal” parameters as switches to allow variant cases to be easily set up from a common
“source” file. Literal parameters used include:
TypeScenario – is set to ‘Design’, ‘DefectD1’, ‘DefectD2’, ‘DefectD3’, ‘DefectD4’, or
‘BhEros’9; and
TypeGeosphere – is set to ‘AerobicFractured’, ‘AerobicPorous’,‘Anaerobic Fractured’, or
‘AnaerobicPorous’ to account for uncertainty in the nature of the oxidising/reducing
conditions and the geosphere flow
9
Although the erosion of the cover above the borehole is included in the Design Scenario, for the purposes of the
AMBER modelling it is account for using a different TypeScenario literal.
71
6
PRESENTATION AND ANALYSIS OF RESULTS
Before presenting and analysing the results of the PCSA, it is important to summarise the
main assumptions that have been adopted. This ensures that the reader is aware of these
assumptions and can review their appropriateness when applying the PCSA and its results
to a specific disposal system.
The main assumptions are summarised in the first column of Table 17 (those appearing in
italics relate to parameters that are site-specific). These assumptions have been identified
by reviewing each step of the approach used in the PCSA (i.e. the specification of the
assessment context (Section 2), the description of the disposal system (Section 3), the
development and justification of the scenarios (Section 4), and the formulation and
implementation of models (Section 5)). Where appropriate, each assumption has been
classified (in the second and third columns) as to whether it is considered to be
conservative or realistic, consistent with the definitions of these terms provided in Section
2.6.2. The sections of this report that provide the justification for each assumption are
listed in the fourth column.
TABLE 19. KEY ASSUMPTIONS MADE IN THE PCSA
Assumption
1. Narrow diameter borehole
(up to 50 cm) and so small
diameter sources (up to
15 mm)
2. Disused sealed sources
3. Only consider post-closure
issues
4. Exclude deliberate human
intrusion
5. Depth of cover 50 m
6. Assume that the derived
reference activity values are
total values applicable to an
entire site
7. Only consider radiological
impacts on humans
8. Regulatory framework and
associated end points
9. No explicit consideration of
radiolysis, criticality and
thermal effects
Conservative
Assessment Context
N/A
Realistic
Justification
N/A
Sections 1.3 and 2.1
N/A
N/A
N/A
N/A
Sections 1.3, 2.1 and
2.1
Sections 1.3 and 2.1
N/A
N/A
Section 1.3
N/A
Yes
N/A
-
Sections 1.3 and 2.1
Section 2.1
N/A
N/A
Section 2.1
N/A
N/A
Sections 2.3 and 2.4
No
No
Such effects are
considered to be
insignificant for the
72
Assumption
Conservative
Realistic
-
Yes
Justification
typical inventories to
be disposed
Section 2.6
N/A
Yes
N/A
Section 2.6
Section 2.6
Yes
Section 3.1.1 and
Appendix A
-
Yes
Section 3.1.2
N/A
N/A
Section 3.1
N/A
Yes
Section 3.1.2
5. Borehole located in saturated
zone s
6. Absence of geological
complexity and variability
can be averaged
7. Water abstraction borehole
as GBI (drilled at end of
institutional control period)
8. Geological stability (tectonic
and seismic)
9. No natural resources
requiring excavation
10. Flux and travel time through
geosphere and distance to
GBI
11. Sorption coefficients
N/A
Yes
Section 3.1.3
N/A
Yes
Section 3.2
Yes
-
Section 3.2
N/A
N/A
Section 3.2
N/A
N/A
Section 3.2
N/A
N/A
Sections 3.2.1 and
3.2.2
N/A
N/A
12. Climatic conditions
13. Soils capable of supporting
crops
14. Subdued relief
15. Limited geomorphological
activity (e.g. no coastal
processes)
N/A
N/A
N/A
N/A
Sections 3.3.1 and
3.3.2
Section 3.3
Section 3.3
N/A
N/A
N/A
N/A
Section 3.3
Section 3.3
1. Identified scenarios
adequately illustrate the
range of future behaviours
-
Yes
Section 4
10. Borehole operated only for
one year and then closed
11. Period of institutional control
12. No cut-off time for
calculation of dose
1. The radionuclides are
representative of those that
can be found in sealed
sources
2. Sources have been
appropriately conditioned
prior to disposal
3. Using borehole disposal
concept design and materials
broadly similar to that
defined in Section 3.1
4. 43 waste packages
System Description
-
Scenarios
73
Assumption
and states of the disposal
system
2. Unexpected geological
conditions and environmental
change scenarios are
adequately covered by the
other scenarios and
associated variant
calculations
Conservative
Realistic
Justification
-
Yes
Sections 4.4 and 4.5
-
Yes
Section 5.3
Yes
(for liquid release of
certain solubility
limited
radionuclides)
No
Yes
(for most
radionuclides and
releases)
Section 5.3
-
Section 5.3
Some
Some
Section 5.5
Models
1. Use of the compartment
modelling approach is
appropriate for the problem
2. Linear relationship between
activity and dose
3. Activity levels derived by
considering radionuclides
independently
4. Assume that the data used are
appropriate
6.1 RESULTS FOR THE REFERENCE CALCULATIONS
6.1.1 Design Scenario
Table 20. Total Peak Dose for Aerobic Fractured (Design Scenario)
Radionuclide
Dose (Sv/y)
Co-60
Sr-90
I-129
Cs-137
<1E-15
4.3E-14
3.0E-3
<1E-15
Ra-226
Chain
9.3E-6
Am-241
Chain
1.8E-8
Cf-252
Chain
1.5E-13
Total
3.0E-3
The assessment results for liquid releases for the design scenario are presented in Table 20
and Figure 17 in terms of the peak dose an observer is exposed to at the end of the
institutional control period. From the table, it could be seen that the total dose resulting
from all the radionuclides and their chains is 3.0E-3Sv/y which is higher than the dose
criterion of 0.3mSv/y. From the table and the Figure, the dose from all the radionuclides is
insignificant except that from I-129 (3.0E-3Sv/y) hence the need for different disposal
option for this radionuclide.
The total peak dose from the daughters of various radionuclides is shown in Table 21. The
dose for various times has also been presented in Figure 17. It can be seen from the table
that, the contribution from Cf-252 is negligible (less than 1E-15 Sv/yr) but some of the
daughters contribute appreciable doses.
74
Table 21. Total Peak dose from radionuclides and their daughters for Aerobic Fractured
(Design Scenario)
Chain
Ra-226
Members
Ra-226
Pb-210
Po-210
Am-241
Np-237
U-233
Pa-233
Th-229
Cf-252
Cm-248
Pu-244
Pu-240
U-236
Th-232
Ra-228
Am-241
Cf-252
Peak Dose (Sv/yr)
2.6E-7
3.3E-6
5.7E-6
1.2E-11
1.9E-9
7.9E-10
1.9E-11
1.5E-8
<1E-15
2.4E-15
6.8E-14
7.0E-14
1.3E-14
<1E-15
<1E-15
D_Tot_Chain[Farmer]
1
0.1
D_Tot_Chain (Sv y-1)
0.01
0.001
Co60
0.0001
Sr90
1E-05
I129
1E-06
Cs137
Ra226
1E-07
Am241
1E-08
Cf252
1E-09
1E-10
1
100
10000
1000000
Time (Years)
Figure 17.
Scenario)
Calculated Dose from the Radionuclides for Aerobic Fractured (Design
75
Table 22. Total Peak Dose Due to Erosion of Closure zone
Radionuclide
Dose (Sv/y)
Co-60
0
Sr-90
0
I-129
8.5E-1
Cs-137
0
Ra-226
Chain
2.7E-11
Am-241
Chain
5.3E-4
Cf-252
Chain
2.7E-7
Total peak dose for solid releases after the waste have been uncovered by erosion are
provided in Table 22. They show that all radionuclides, other than I-129 and those with
long-lived daughters (i.e. Ra-226, Am-241 and Cf-252) decay before the waste is
uncovered. This is also depicted in Figure 18. But for I-129, the dose that a site dweller
will be exposed to from any of the radionuclides is less than 0.3mSv/y.
D_Tot_Chain[SiteDwellerErosion]
D_Tot_Chain (Sv y-1)
1
0.1
0.01
0.001
Co60
0.0001
Sr90
0.00001
I129
Cs137
0.000001
Ra226
0.0000001
Am241
1E-08
Cf252
1E-09
1E-10
5000
50000
500000
Time (Years)
Figure 18. Calculated Dose from the Radionuclides for Aerobic Fractured (Design
Scenario, Solid Releases to Site Dweller)
6.1.2
Defect Scenario
Table 23. Total Peak Dose for Aerobic Fractured (D4 Scenario)
76
Radionuclide
Co-60
Dose (Sv/y)
Sr-90
0
I-129
1.4E-12
5.8E-1
Cs-137
4.1E-14
Ra-226
Chain
1.2E-5
Am-241
Chain
3.5E-8
Cf-252
Chain
3.0E-13
In this case a particular capsule is failed and this failed capsule is in a failed container. If
the failed capsule contains Co-60, then the observer will not be exposed to any dose.
However, if it contains Sr-90, then the observer will be exposed to a peak dose of 1.4E-12
Svy-1 and this applies to all radionuclides shown in Table 23 and Figure 18. I-129 is of
much concern for this scenario as the peak dose, 5.8E-1 Sv/y , exceeds the acceptance dose
of 0.3mSv/y.
D_Tot_Chain[Farmer]
D_Tot_Chain (Sv y-1)
1
0.1
0.01
0.001
Co60
0.0001
Sr90
0.00001
I129
0.000001
Cs137
0.0000001
Ra226
1E-08
Am241
1E-09
Cf252
1E-10
1
10
100
1000
10000
100000 1000000 10000000
Time (Years)
Figure 18. Calculated Dose from the Radionuclides for Aerobic Fractured (D4 Scenario)
Table 24. Total Peak Dose for Aerobic Fractured (D3 Scenario)
Radionuclide
Dose (Sv/y)
Co-60
0
Sr-90
I-129
Cs-137
1.7E-11
1.4E+0
4.7E-13
Ra-226
Chain
2.0E-5
Am-241
Chain
3.5E-8
Cf-252
Chain
3.0E-13
Total
1.4E+0
The dose that an observer will be exposed to, in the event that the cement grout in
disposal/disturbed zone is degraded/incomplete (D3 Scenario) is presented in Table 24 and
Figure 19.
Apart from I-129, the total peak dose from all the radionuclides for this scenario is below
the acceptance dose of 0.3mSv/y.
77
D_Tot_Chain[Farmer]
D_Tot_Chain (Sv y-1)
10
1
0.1
0.01
Co60
0.001
Sr90
0.0001
I129
0.00001
Cs137
0.000001
Ra226
0.0000001
Am241
Cf252
1E-08
1E-09
1E-10
1
10
100
1000
10000
100000 1000000 10000000
Time (Years)
Figure 18. Calculated Dose from the Radionuclides for Aerobic Fractured (D3 Scenario)
6.2
RESULTS FOR VARIANT CALCULATIONS
In order to investigate the sensitivity of the results presented in Section 6.1, to conceptual
model and data assumptions, 3 variant cases were identified and associated calculations
undertaken. The cases considered:

Aerobic Porous geosphere

Anaerobic Porous geosphere

Anaerobic Fractured geosphere
The results for these cases are presented in Tables 25 to 27 and Figures 19 to 21. For all the
cases, the total peak dose that an observer is exposed to are below 0.3mSv/y except the
case of Aerobic Porous geosphere in which the total is 3.0E-3 Sv/y.
78
Table 25. Total Peak Dose for Aerobic Porous (Design Scenario)
Radionuclide
Co-60
Dose (Sv/y)
Sr-90
0
I-129
9.3E-14
3.0E-3
Cs-137
2.7E-15
Ra-226
Chain
1.7E-4
Am-241
Chain
2.7E-5
Cf-252
Chain
4.8E-9
Total
3.0E-3
D_Tot_Chain[Farmer]
D_Tot_Chain (Sv y-1)
1
0.1
0.01
0.001
Co60
0.0001
Sr90
0.00001
I129
Cs137
0.000001
Ra226
0.0000001
Am241
1E-08
Cf252
1E-09
1E-10
1
10
100
1000
10000
100000
1000000 10000000
Time (Years)
Figure 19. Calculated Dose from the Radionuclides for Aerobic Porous (Design Scenario)
79
Table 26. Total Peak Dose for Anaerobic Porous (Design Scenario)
Radionuclide
Co-60
Dose(Sv/y)
Sr-90
0
I-129
0
1.8E-3
Cs-137
0
Ra-226
Chain
5.8E-6
Am-241
Chain
2.9E-5
Cf-252
Chain
4.7E-9
Total
1.8E-3
D_Tot_Chain[Farmer]
D_Tot_Chain (Sv y-1)
1
0.1
0.01
0.001
Co60
0.0001
Sr90
I129
0.00001
Cs137
0.000001
Ra226
0.0000001
Am241
1E-08
Cf252
1E-09
1E-10
1
10
100
1000
10000
100000 1000000 10000000
Time (Years)
Figure 20. Calculated Dose from the Radionuclides for Anaerobic Porous (Design
Scenario)
80
Table 27. Total Peak Dose for Anaerobic Fractured (Design Scenario)
Radionuclide
Co-60
Sr-90
<1E-15
<1E-15
I-129
1.8E-3
Cs-137
<1E-15
Ra-226
Chain
3.2E-7
Am-241
Chain
1.8E-8
Cf-252
Chain
1.5E-13
Total
1.8E-3
Dose (Sv/y)
D_Tot_Chain[Farmer]
D_Tot_Chain (Sv y-1)
1
0.1
0.01
0.001
Co60
0.0001
Sr90
I129
0.00001
Cs137
0.000001
Ra226
0.0000001
Am241
1E-08
Cf252
1E-09
1E-10
1
10
100
1000
10000
100000 1000000 10000000
Time (Years)
Figure 21. Calculated Dose from the Radionuclides for Anaerobic Fractured (Design
Scenario)
6.3
ANALYSIS OF UNCERTAINTIES
When undertaking a long-term safety assessment of a radioactive waste disposal system, it
is important to be aware of and to manage, as far as possible, the various sources of
uncertainty that arise. In addition, appropriate steps should be taken to build confidence in
the assessment and its results. Various measures have been implemented as part of the
current assessment to address uncertainties and build confidence.
Uncertainties can be considered to arise from three sources (IAEA, 1993).
81
First there is uncertainty in the evolution of the disposal system over the timescales of
interest (scenario uncertainty). This has been accounted for in the current assessment by
considering five scenarios, two of which have been evaluated quantitatively (Design and
Defect Scenarios). The development and justification of these scenarios is discussed in
Section 4. For the given disposal system, the range in associated activity levels for the two
scenarios assessed quantitatively is generally small (much less than an order of
magnitude). However, Defect Scenario D4 (which involves a failed waste capsule being
within a failed disposal container) does result in a total peak dose of 5.8E-1Sv/y for I-129
compared with values in excess of 5.77E-1Sv/y for the Design Scenario.
The second source of uncertainty is uncertainty in the conceptual, mathematical and
computer models used to simulate the behaviour and evolution of the disposal system (e.g.
owing to the inability of models to represent the system completely, approximations used
in solving the model equations, and coding errors) (model uncertainty). Various quality
assurance checks have been undertaken to ensure that the mathematical model and data
specified in Appendix C and D have been correctly implemented in the AMBER software
tool and, as discussed in Section 5.6, independent verification tests for AMBER have been
undertaken. Different concepts for the release of radionuclides from the near field and the
use of abstracted water have been considered in Section 4.2.1, indicating that differences
can arise compared to the reference assumptions.
The PCSA considers that the GBI to be a water abstraction borehole and erosion of the
closure zone (see Section 4.2). Alternative interfaces could be considered in future
iterations that might result in the accumulation of certain long-lived radionuclides (e.g.
groundwater discharge into lake sediment). This sediment could subsequently be
uncovered resulting in the exposure of humans to the accumulated radionuclides.
The third source of uncertainty is uncertainty in the data and parameters used as inputs in
the modelling (data and parameter uncertainty). This first iteration of the PCSA made
use of regional geosphere data as site specific data is not available yet. Such uncertainties
can be assessed through deterministic and/or probabilistic sensitivity analysis if resources
allow. The range of different geospheres considered in the current assessment allows an
initial assessment of the impact of different parameter values (e.g. different corrosion and
degradation rates, sorption coefficients, and hydraulic conductivities). In addition,
deterministic calculations have been reported in Section 6.2 that illustrate the sensitivity of
the results to different parameter values. Differences of high orders of magnitude are
observed for certain radionuclides for some parameters (e.g. geosphere pathlength and
sorption coefficient). But for other parameters, the differences are significantly less than an
order of magnitude.
In addition to the above sources of uncertainty, a further type of uncertainty, subjective
uncertainty (uncertainty due to reliance on expert judgement), is also linked with the above
sources of uncertainty (IAEA, 2004). In common with many other assessments, expert
judgement has been used at many stages during the current assessment due to a variety of
reasons such as a lack of knowledge concerning current and future conditions, conceptual
models and data/parameter values (and distributions). Where such judgements have been
made in the current assessment, they have been documented and, as far as practicable,
justified – see for example Section 4 for the scenario development and justification process
and Appendix D for the data values.
82
6.4
BUILDING OF CONFIDENCE
Thorough discussions within various international fora (such as the Nuclear Energy
Agency (NEA, 2004a, 2004b) and the (IAEA 2003, 2004) it is becoming increasingly
recognised that building confidence in the long-term safety of a radioactive waste
repository is an increasingly important issue. To undertake a safety assessment and present
the results is not sufficient. Confidence needs to be built in the safety assessment and its
results. There is also a need to have confidence in other aspects of the long-term safety of
the repository in order to build confidence to the satisfaction of all stakeholders (i.e.
regulators, the public, wider scientific community, political decision makers etc.). In
particular, confidence in the long-term safety needs to be promoted and communicated
through a more broadly based ‘safety case’ IAEA, 2003. The safety case puts the findings
of the safety assessment into a broader context with other factors and considerations that
are relevant to the decision making process and are important for the stakeholders
involved.
Given that the focus of the current document is the safety assessment rather than the
broader based safety case, the emphasis of this sub-section is on measures that have been
taken to building confidence in the safety assessment and its results. Confidence in the
safety assessment should be established at two levels IAEA, 2003. The first level involves
establishing confidence within each stage of the safety assessment process (i.e. assessment
context, system description, development and justification of scenarios, formulation and
implementation of models, analysis of the results, and review, modification and subsequent
iterations). The second level involves gaining an overall confidence, which involves
gaining confidence in the overall safety assessment methodology, safety assessment
approach and the safety assessment findings through the use of a range of techniques
IAEA, 2003. The measures undertaken within the current assessment to building
confidence at these two stages are summarised in Table 28 and Table 29.
83
TABLE 3: MEASURES TAKEN TO BUILD CONFIDENCE IN EACH STAGE OF THE SAFETY ASSESSMENT
Stage of the
Assessment
Specification of
assessment
context
Description of
the system
Development
and justification
of scenarios
Formulation and
implementation
of models
Confidence Building Measures taken in the PCSA


The assessment context for the PCSA of the borehole disposal concept is explained and justified in detail in Section 2.
Each of the components of the context (purpose and scope, regulatory framework, end points, philosophy, and timeframes) is discussed in turn.


Different geospheres (aerobic, anaerobic, porous and fractured) are considered.
The approach used is consistent with that used in a previous IAEA study to derive generic activity levels IAEA, 2003and a previous generic safety
assessment of a borehole disposal concept, Little et. al., 2004
Representative information has been taken from a range of relevant sources documented in Section 3 and Appendix D.
A set of five scenarios has been developed and justified in Section 4.
It is considered to be a credible and comprehensive set and to have been developed in a systematic, transparent and traceable manner using an
international panel of experts with differing fields of expertise.
The approach and screening criteria used to exclude or include scenarios has been justified and documented.
The development of conceptual models, consistent with the assessment context, the disposal systems and with the scenarios to be investigated, has been
undertaken in a systematic manner consistent with best international practice (Section 5).
The mathematical models used are consistent with those used in previous assessments such as (IAEA , 2003, 2004), Little et. al., 2004 and data are
derived from a wide range of published and internationally recognised references.
The software tool in which the mathematical models are encoded (AMBER) has been used in a variety of assessments (see for example (IAEA , 2003,
2004), Little et. al., 2004 ,Maul and Robinson, 2002 and Penford et.al., 2003 ) The verification of the time dependent solution method used by AMBER
solution is discussed in Robinson et. al., 2006. The implementation of the models in the software has been audited.
Consideration has been given to the use of alternative conceptual models and data.
The results obtained for the wide range of disposal systems and scenarios have been presented in Sections 3 and 4 and compared against the relevant
regulatory criteria.
Both referenceand variant calculations have been considered.
Consideration has been given in Section 6.3 to the various sources of uncertainty (scenario, model and data/parameter).







Analysis of the
results




84
TABLE 29: MEASURES TAKEN TO BUILD CONFIDENCE IN THE OVERALL SAFETY ASSESSMENT
Confidence Building Measures
Application to the PCSA
Use of a systematic approach
An approach based upon the internationally recognised ISAM Safety Assessment Approach IAEA, 2004a,
2004b has been used. The approach allows the PCSA and its associated assumptions to be documented in a
clear manner.
Peer review
Yet to be peer reviewed.
Quality assurance
The assessment has been carried out under Quintessa’s Quality Management System, which is compliant with
the ISO 9001:2000 standard IOS, 2000
Verification, calibration and, if
possible, validation of models
The verification of the time dependent solution method used by AMBER solution is discussed in Robinson et.
al., 2006. Due long timescales, validation of the models is not considered to be possible.
Consideration of relevant
analogues
No consideration has been given to this issue in the PCSA.
Involvement of stakeholders
IAEA Experts (Richard Little, Mathew W.Kozak ,and Jacobus J.V. Blerk) have been involved in the
development of the approach followed and the identification and justification of the scenarios for assessment.
Consideration and treatment of
uncertainties
This is discussed in Section 6.3.
Presentation of the assessment
and its results
The results are presented in Sections 6.1 and 6.2. The other components of the assessment are discussed in
Sections 2 to 5 and the supporting appendices.
85
7
CONCLUSIONS

This is an initial first iteration of a Ghana-specific post-closure safety
assessment – it will be refined in light of further iterations.

It has shown that the IAEA’s GSA is a useful starting point for the
development of country-specific assessment.

The assessment indicates that Ghana’s current inventory of disused sealed
sources that cannot be repatriated appear to be capable of being safely
disposed using the borehole disposal concept. Some have key caveats are:
o An alternative disposal option needs to be found for the liquid H-3
waste
o The I-129 needs to be further investigated and might not be suitable
for disposal in the BDF
o Further characterisation is required of the sources, geosphere and
biosphere.

In terms of source characterisation, assumptions have been adopted relating
to the dimensions and activity levels of the sources (Section 3.1.1). These
need to be reviewed/revised in light of on-going source characterisation work.
The source dimensions impact on the size of capsule that can be used.

In terms of geosphere characterisation, assumptions have been adopted
relating to the geosphere (Section 3.2). These need to be reviewed /revised in
light of the proposed geosphere characterisation work – two boreholes are to
be sunk in the next 6 months. It will be particularly helpful to obtain data on
the nature of groundwater flows (fracture vs porous), the hydraulic
parameters (hydraulic conductivity, gradient, porosity), salinity and Eh
conditions.

In terms of biosphere characterisation, more site-specific information could
be collected. However, it is noted that biosphere characteristics are likely to
change with time and so the use of the generic parameters (that are chosen to
maximise impacts) used in the current assessment might be adequate.

In terms of undertaking the next iteration of the assessment, work would need
to be done on:
o FEP screening;
o more detailed groundwater flow modelling using the results from the
site investigation boreholes to inform a groundwater modelling code;
86
o reviewing the corrosion and cement degradation models in light of the
geochemical and groundwater flow data to be obtained from the site
investigation boreholes;
o extending the range of calculations performed (e.g., additional
sensitivity cases, consideration of non-human biota, consideration of
non-rads, etc.).
87
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93
APPENDIX A: SCREENING OF SOURCES
The inventory of disused sources for potential disposal in the BDF is given in Table
A.1 (based on the list given in Table 3 in Section 3.1).Table A.1 also provides the
half-life for each radionuclide. For the purposes of these screening calculations, the
activity associated with each type of source is assumed to be distributed evenly over
the number of capsules to be used for that source type.
TABLE A.1. SOURCES CONSIDERED
SCREENING CALCULATIONS
Radionuclide
No. of
No. of
Sources containers
IN
THE
Total
Initial
Inventory
(Bq)
Half-life
(y)
Na-22
4
1
3.70E+06
2.60E+00
P-32
4
1
1.18E+09
3.91E-02
Ca-45
3
1
1.85E+08
4.46E-01
Co-57
3
1
1.11E+08
7.42E-01
Fe-59
2
1
2.22E+10
1.22E-01
Co-60
2
1
1.75E+06
5.27E+00
Co-60
2
1
4.90E+05
5.27E+00
Co-60
1
1
2.78E+14
5.27E+00
Co-60
1
1
1.85E+14
5.27E+00
Co-60
1
5
2.22E+14
5.27E+00
Zn-65
1
1
3.70E+08
6.68E-01
Sr-89
1
1
4.77E+09
1.38E-01
Sr-90
33
4
1.25E+10
2.91E+01
Cd-109
6
1
6.66E+08
1.27E+00
In-113m
12
1
2.22E+09
1.89E-04
I-129
1
1
4.25E+10
1.57E+07
I-131
2
1
6.21E+09
2.20E-02
DECAY-STORAGE
94
Cs-137
30
5
5.66E+12
3.00E+01
Cs-137
2
1
4.09E+06
3.00E+01
Cs-137
3
1
3.70E+11
3.00E+01
Cs-137
1
1
3.00E+07
3.00E+01
Ir-192
1
1
2.26E+12
2.03E-01
Tl-204
2
1
7.40E+05
3.78E+00
Ra-226
19
4
7.03E+09
1.60E+03
Am-241
3
1
1.85E+12
4.32E+02
Am-241
1
1
1.80E+09
4.32E+02
Am-241
105
11
3.50E+07
4.32E+02
Am-241
1
1
1.67E+09
4.32E+02
Cf-252
2
1
2.22E+10
2.64E+00
Two screening calculations have been undertaken:


A.1
a calculation to determine which sources are suitable for decay storage rather
than borehole disposal (Appendix A.1); and
a calculation to determine which sources to be disposed in the borehole do
not need to be assessed in detail (Appendix A.2).
DECAY-STORAGE SCREENING
A number of sources in Table A.1 contain radionuclides with half lives of much less
than a year and so could potentially be decay stored rather than disposed in the BDF.
In order to identify suitable sources for decay storage, a spreadsheet has been
developed to allow the calculation of doses associated with direct exposure via
ingestion, inhalation and external irradiation to a source. The results are reproduced
in Table A.2. The initial activity (i.e., the activity at manufacture) of each source
type given in Table A.1 has been decay-corrected to account for 10 years of use and
storage prior to disposal. Sources that result in a dose of less than 1 µSv y-1 are
considered suitable for decay storage and do not need to be disposed in the BDF.
These are the P-32, Ca-45, Fe-59, Sr-89, In-113m, I-131 and Ir-192 sources.
95
Table A.2: Decay Storage Screening Calculations
Radionuclide
Na-22
No. of
Sources
4
P-32
4
S-35*
5
Ca-45
3
Co-57
3
Total
Initial
Inv (Bq)
Half-life
(y)
3.70E+06
1.18E+09
Decayed
inv per
container
(Bq)
Ingestion
Dose
(Sv)
Inhalation
Dose (Sv)
External
Dose
(Sv)
MeV
Ingestion
(Sv/Bq)
Inhalation
(Sv/Bq)
External
(Sv/h per
Bq)
2.60E+00
2.19E+00
3.2E-09
1.3E-09
3.1E-13
2.6E+05
8.2E-04
3.3E-04
3.2E-06
3.91E-02
1.07E-03
2.4E-09
3.4E-09
1.5E-16
1.2E-68
2.9E-77
4.1E-77
7.2E-83
2.39E-01
2.09E-06
7.7E-10
1.9E-09
2.9E-19
0.0E+00
0.0E+00
0.0E+00
0.0E+00
1.85E+08
4.46E-01
8.80E-06
7.1E-10
2.7E-09
1.2E-18
3.3E+01
2.3E-08
8.9E-08
1.6E-15
1.11E+08
7.42E-01
1.20E-01
2.1E-10
5.5E-10
1.7E-14
9.7E+03
2.0E-06
5.4E-06
6.5E-09
Fe-59
2
2.22E+10
1.22E-01
1.19E+00
1.8E-09
3.7E-09
1.7E-13
4.7E-15
8.5E-24
1.7E-23
3.1E-26
Co-60
2
1.75E+06
5.27E+00
2.50E+00
3.4E-09
1.0E-08
3.5E-13
4.7E+05
1.6E-03
4.7E-03
6.6E-06
Co-60
2
4.90E+05
5.27E+00
2.50E+00
3.4E-09
1.0E-08
3.5E-13
1.3E+05
4.5E-04
1.3E-03
1.8E-06
Co-60
1
2.78E+14
5.27E+00
2.50E+00
3.4E-09
1.0E-08
3.5E-13
7.5E+13
2.5E+05
7.5E+05
1.0E+03
Co-60
1
1.85E+14
5.27E+00
2.50E+00
3.4E-09
1.0E-08
3.5E-13
5.0E+13
1.7E+05
5.0E+05
7.0E+02
Co-60
1
2.22E+14
5.27E+00
2.50E+00
3.4E-09
1.0E-08
3.5E-13
1.2E+13
4.1E+04
1.2E+05
1.7E+02
Zn-65
1
3.70E+08
6.68E-01
5.81E-01
3.9E-09
1.6E-09
8.1E-14
1.2E+04
4.5E-05
1.8E-05
3.8E-08
Sr-89
1
8.45E-05
2.6E-09
6.1E-09
1.2E-17
7.3E-13
1.9E-21
4.5E-21
3.5E-28
33
4.77E+09
1.25E+10
1.38E-01
Sr-90
2.91E+01
2.00E-03
3.1E-08
3.8E-08
2.8E-16
2.5E+09
7.6E+01
9.4E+01
2.8E-05
Cd-109
6
6.66E+08
1.27E+00
3.18E-03
2.0E-09
8.1E-09
4.5E-16
2.8E+06
5.7E-03
2.3E-02
5.1E-08
In-113m
12
2.22E+09
1.89E-04
2.52E-01
2.8E-11
2.0E-11
3.5E-14
0.0E+00
0.0E+00
0.0E+00
0.0E+00
I-129
1
4.25E+10
1.57E+07
1.27E-06
1.1E-07
3.6E-08
1.8E-19
4.2E+10
4.7E+03
1.5E+03
I-131
2
6.21E+09
2.20E-02
3.79E-01
2.2E-08
7.4E-09
5.3E-14
9.1E-128
2.0E-135
6.8E-136
3.0E-07
1.9E139
Age of source
(years)
Duration of
exposure (hours)
Dose constraint
(Sv/y)
10
40
1.00E-06
96
Cs-137
30
5.66E+12
3.00E+01
5.60E-01
1.3E-08
3.7E-08
7.8E-14
9.0E+11
1.2E+04
3.3E+04
2.8E+00
Cs-137
2
4.09E+06
3.00E+01
5.60E-01
1.3E-08
3.7E-08
7.8E-14
3.2E+06
4.2E-02
1.2E-01
1.0E-05
Cs-137
3
Cs-137
Ir-192
Tl-204
Ra-226
Am-241
Am-241
Am-241
Am-241
Cf-252
1
1
2
19
3
1
105
1
2
3.70E+11
3.00E+07
2.26E+12
7.40E+05
7.03E+09
1.85E+12
1.80E+09
3.50E+07
1.67E+09
2.22E+10
3.00E+01
3.00E+01
2.03E-01
3.78E+00
1.60E+03
4.32E+02
4.32E+02
4.32E+02
4.32E+02
2.64E+00
5.60E-01
5.60E-01
8.11E-01
1.05E-03
1.70E+00
2.13E-02
2.13E-02
2.13E-02
2.13E-02
6.12E+00
1.3E-08
1.3E-08
1.4E-09
1.2E-09
2.8E-07
2.0E-07
2.0E-07
2.0E-07
2.0E-07
9.0E-08
3.7E-08
3.7E-08
6.6E-09
3.9E-10
3.5E-06
4.2E-05
4.2E-05
4.2E-05
4.2E-05
2.0E-05
7.8E-14
7.8E-14
1.1E-13
1.5E-16
2.4E-13
3.0E-15
3.0E-15
3.0E-15
3.0E-15
8.6E-13
2.9E+11
2.4E+07
3.3E-03
1.2E+05
1.7E+09
1.8E+12
1.8E+09
3.1E+06
1.6E+09
1.6E+09
3.8E+03
3.1E-01
4.7E-12
1.4E-04
4.9E+02
3.6E+05
3.5E+02
6.3E-01
3.3E+02
1.4E+02
1.1E+04
8.8E-01
2.2E-11
4.6E-05
6.1E+03
7.6E+07
7.4E+04
1.3E+02
6.9E+04
3.2E+04
9.2E-01
7.5E-05
1.5E-14
7.0E-10
1.7E-02
2.2E-01
2.1E-04
3.7E-07
2.0E-04
5.5E-02
Notes
1. Short-lived daughters with a half life of less than 25 days are assumed to be in secular equilibrium with their parent and included in
the parent’s dose coefficient. A list of short-lived daughters is given in Table A.3.
2. Data taken from ICRP (1996) for adults.
3. Data taken from ICRP (1996) for adults, adopting the recommended default absorption class, where no recommendation is made, then
the most conservative (highest) dose coefficient is adopted from the range of absorption classes reported.
4. Dose factor for point source at 1 m calculated by multiplying mean gamma energy in MeV by 1.4E-13 Sv/h per Bq/MeV (Smith et
al., 1988). Emissions data are taken from ICRP (1983) and Browne and Firestone (1988). Photons with individual energies below 50
keV have not been included because the equation used to calculate the dose coefficient from a point source substantially over-estimates
the dose rate below this value, and the contribution to effective dose equivalent, given the existence of other exposure pathways, would
in any event be very small.
5. Dose calculated assuming exposure duration of 40 hours.
97
References
ICRP (International Comission on Radiological Protection). 1996. Age-dependent
doses to members of the public from intake of radionucldes: Part 5. Compilation of
ingestion and inhalation dose coefficients.Annals of the ICRP 26(1), ICRP
Publication 72, Pergamon Press. Oxford, UK.
Smith G M, Fearn H S, Smith K R, Davis J P and Klos R (1988). Assessment of the
radiological impact of disposal of solid radioactive waste at Drigg. National
Radiological Protection Board, NRPB-M148, Chilton, UK.
ICRP (1983).
Radionuclide Transformations Energy and Intensity of
Emissions.International Commission on Radiological Protection, ICRP Publication
38. Pergamon Press, Oxford.
Browne and Firestone (1988). Table of the Radioactive Isotopes.J Wiley and Sons.
TABLE A.3. SHORT-LIVED DAUGHTERS WITH HALF-LIVES OF LESS
THAN 25 DAYS ASSUMED TO BE IN SECULAR EQUILIBRIUM WITH
THEIR PARENTS
Parent
Sr-90
Cs-137
Pb-210
Ra-226
Th-229
Short Lived Daughters
 Y-90
 (branching ratio 0.946) Ba-137m
 Bi-210
 Rn-222 Po-218 (branching ratio 0.9998) Pb-214  Bi-214  (branching ratio 0.9998) Po-214
 (branching ratio 0.0002) At-218  Bi-214  (branching ratio 0.9998) Po-214
 Ra-225 Ac-225 Fr-221 At-217 Bi-213 (branching ratio 0.9784) Po-213 Pb-209
 (branching ratio 0.0216) Tl-209 Pb-209
A.2
ASSESSMENT SCREENING
Of the remaining sources that are to be disposed in the BDF, it is possible to
undertake a further screening calculation to identify those sources which contain
radionuclides that, due to their half-life, maximum activity, or radiotoxicity, will not
result in significant post-closure impacts and so do not need to be assessed in detail.
A dose constraint of 0.3 mSv y-1 is applied for these calculations (consistent with the
constraint for disposal given in Section 2.3).
Institutional control periods are often taken into consideration such that there is a
period within which exposures are assumed not to occur. For this PCSA an
institutional control period of 50 years has been adopted (Section 2.6), within which
exposures are considered not to occur. However, for the purpose of these screening
98
calculations a shorter, more conservative institutional control period of 30 years is
assumed.
Doses associated with direct exposure via ingestion, inhalation and external
irradiation to a single disused source following a 30 year decay period (to present the
institutional control period) are calculated using the same dose coefficients as used in
the decay-storage screening calculations (Table A.2)
The screening calculation assumes that a human is directly exposed to a single sealed
source following the end of the institutional control period. Exposure through
ingestion, inhalation and external irradiation is considered. The resulting doses are
given in Table A.4.
TABLE A.4.–ASSESSMENT SCREENING: DOSES ASSOCIATED WITH
DIRECT EXPOSURE TO A SEALED SOURCE
Radionuclide No. of
No. of
in source
Sources containers
Na-22
4
1
Co-57
3
1
Co-60
2
1
Co-60
2
1
Co-60
1
1
Co-60
1
1
Co-60
1
5
Zn-65
1
1
Sr-90
33
4
Cd-109
6
1
I-129
1
1
Cs-137
30
5
Cs-137
2
1
Cs-137
3
1
Cs-137
1
1
Tl-204
2
1
Ra-226
19
4
Am-241
3
1
Am-241
1
1
Am-241
105
11
Am-241
1
1
Cf-252
2
1
Activity after
30 y per
container
(Bq)
1.2E+03
7.5E-05
3.4E+04
9.5E+03
5.4E+12
3.6E+12
8.6E+11
1.1E-05
1.5E+09
5.2E+01
4.2E+10
5.7E+11
2.0E+06
1.9E+11
1.5E+07
3.0E+03
1.7E+09
1.8E+12
1.7E+09
3.0E+06
1.6E+09
8.4E+06
Ingestion
External
Dose
Inhalation
Dose
(Sv)
Dose (Sv)
(Sv)
4.0E-06
1.6E-06
1.5E-08
1.6E-14
4.1E-14
5.0E-17
1.2E-04
3.4E-04
4.7E-07
3.2E-05
9.5E-05
1.3E-07
1.8E+04 5.4E+04 7.5E+01
1.2E+04 3.6E+04 5.0E+01
2.9E+03 8.6E+03 1.2E+01
4.4E-14
1.8E-14
3.6E-17
4.7E+01 5.8E+01 1.7E-05
1.0E-07
4.2E-07
9.2E-13
4.7E+03 1.5E+03 3.0E-07
7.4E+03 2.1E+04 1.8E+00
2.7E-02
7.6E-02
6.4E-06
2.4E+03 6.8E+03 5.8E-01
2.0E-01
5.6E-01
4.7E-05
3.6E-06
1.2E-06
1.8E-11
4.9E+02 6.1E+03 1.7E-02
3.5E+05 7.4E+07 2.1E-01
3.4E+02 7.2E+04 2.0E-04
6.1E-01
1.3E+02 3.6E-07
3.2E+02 6.7E+04 1.9E-04
7.6E-01
1.7E+02 2.9E-04
99
The screening calculations show that the sources containing Na-22, Co-57, Zn-65,
Cd-109 and Tl-204 give rise to doses less than the 0.3 mSv y-1dose constraint and so
can be safely disposed in the BDF and do not need to be assessed in more detailed.
100
APPENDIX B: APPROACH FOR CONCEPTUAL MODEL DEVELOPMENT
Once the scenarios have been developed, their consequences must be analysed. To
allow this, it is necessary to develop a conceptual model of the disposal system, its
environmental setting and the associated release, transport and exposure mechanisms
and media. A conceptual model can be defined as “a set of qualitative assumptions
used to describe a system” [1]. A conceptual model should comprise a description of:
the model’s features, events and processes (FEPs);
the relationships between these FEPs; and
the model’s scope of application in spatial and temporal terms.
The model should have enough detail to allow appropriate mathematical models to
be developed to describe the behaviour of the system and its components.
For the purpose of the current assessment the Interaction Matrix Approach is used to
develop conceptual models in a traceable manner. This approach is based on ideas
developed in BIOMOVS II [2] and subsequently developed and enhanced in a
number of studies such as [3, 4, 5, 6]. The use of the Interaction Matrix allows the
graphical representation of system interactions through the use of formalised
procedures and has the advantage of allowing disposal system components to be
included explicitly in the Interaction Matrix.
The approach starts with a top down approach to dividing the system into constituent
parts. The main components are identified and listed in the leading diagonal
elements (LDEs) of the matrix. The interactions between the LDEs are then noted in
the off-diagonal elements (ODEs). When using the Interaction Matrix approach the
convention is to allocate ODEs in the direction of contaminant migration. In this
way, contaminant migration pathways and the associated exposure pathways and
exposure groups can be traced and translated into the conceptual model. Each
transfer of contaminant from LDE to another LDE via an ODE can be represented by
a mathematical formalism and incorporated into the mathematical model.
As noted above, the first step in developing the Interaction Matrix is to identify the
main components of the disposal system that can be distinguished on the basis of
their chemical and/or physical characteristics. At the top level, the disposal system
can be divided into the near field, geosphere and biosphere. Based on the description
of the disposal system (Section 3) and the scenarios to be assessed (Section 4), the
near field, geosphere and biosphere components listed below can be identified.
B.1
NEAR-FIELD COMPONENTS
Five near-field components can be identified.
Source: The source material, the source container (in which the radioactive source
material is held), and the stainless steel capsule (in which the source container
is assumed to be emplaced). It is conservatively assumed that the source
container will have failed prior to disposal, however the stainless steel capsule
101
is assumed to start to corrode once the disposal container has been breached.
Once the disposal container has been breached, various corrosion mechanisms
(including localised and general corrosion) are assumed to occur and cause the
capsule to be breached (see Table 11).
Containment Barrier: The barrier between the capsule and the disposal container,
which is assumed to be cement grout. Physical and chemical degradation of the
cement grout of barrier is assumed to start once the disposal container has
started to degrade (see Table 11).
Disposal Zone: The stainless steel disposal container, the disposal zone backfill
(cement grout) and the associated borehole casing are considered to comprise
the disposal zone. Whilst the HDPE casing is conservatively assumed to fail on
closure, it is assumed that the stainless steel disposal container remains intact
until breached by corrosion (see Table 11).
Closure Zone: The cement grout backfill, the anti-intrusion barrier and the
uppermost 5 m of native soil or crushed rock are considered to comprise the
closure zone (Figure 4).
Disturbed Zone and Plug: The plug at the bottom of the borehole and the backfill
in the disturbed zone between the casing and the native rock are assumed to be
cement grout (Figure 4).
GEOSPHERE COMPONENTS
At a high level (appropriate for the PCSA), the geosphere can be divided into two
zones.
Unsaturated Zone: comprising the region between the ground surface and the water
table but excluding the rooting zone for major food crops (soil).
Saturated Zone: comprising the region below the water table.
B.3
BIOSPHERE COMPONENTS
Given that the exposure pathways being considered in the Design Scenario are the
domestic and agricultural use of contaminated water by humans and the erosion of
the cover zone above the sources (Section 4.2.1), the biosphere can be sub-divided
into six components.
Humans: who are assumed to be farmers or house dwellers.
Soil: the region in which significant biological activity occurs from the ground
surface to the base of the rooting zone for major food crops.
Atmosphere: the air breathed by humans and fauna, including dust in it.
Crops: the root and green vegetables that are irrigated using contaminated water and
are harvested by humans.
102
Animals: the cattle that are raised by humans and are given contaminated drinking
water.
Elsewhere: Radionuclides can be lost by a number of mechanisms from the
immediate vicinity of the release (e.g. ventilation of the dwelling, groundwater
flow past the abstraction borehole). They are no longer of interest in the
evaluation of individual doses since they are lost to locations where
radionuclide concentrations are lower and the associated doses lower. For the
purpose of this safety assessment and the conceptual model, these locations are
described as being ‘elsewhere’.
B.4
INTERACTIONS BETWEEN COMPONENTS
Based upon expert judgement (gained from previous assessments of the borehole
disposal concept such as [3, 4, 7]) and information from the description of the
disposal system and the scenarios to be assessed, key interactions between the
various disposal system components have been identified that result in the release
and migration of radionuclides through the system and the subsequent exposure of
humans for both the Design and Defect Scenarios. These are shown on Figure 7 for
the case where the disposal zone is saturated.
References for Appendix B
[1]
INTERNATIONAL ATOMIC ENERGY AGENCY, Radioactive Waste
Management Glossary, 2003 Edition, IAEA, Vienna (2003).
[2]
BIOMOVS II, 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 (1996).
[3]
INTERNATIONAL ATOMIC ENERGY AGENCY, Improvement of Safety
Assessment Methodologies for Near Surface Disposal Facilities, Volume II:
Test Cases, IAEA-ISAM-2, IAEA, Vienna (2004).
[4]
LITTLE, R.H., VAN BLERK, J., WALKE, R.C. and BOWDEN, R.A.,
Generic Post-Closure Safety Assessment and Derivation of Activity Limits for
the Borehole Disposal Concept, Quintessa Report QRS-1128A-6 v2.0,
Quintessa Limited, Henley-on-Thames (2004).
[5]
ANDERSSON, J., RIGGARE, P. AND SKAGIUS, K., Project SAFE - Update
of the SFR-1 Safety Assessment Phase 1, SKB Report R-98-43, Swedish
Nuclear Fuel and Waste Management Company, Stockholm (1998).
[6]
BNFL, Drigg Post-Closure Safety Case, British Nuclear Fuels plc, Sellafield
(2002).
[7]
KOZAK, M.W., STENHOUSE, M.J. AND VAN BLERK, J.J., Borehole
Disposal of Spent Sources, Volume II: Preliminary Safety Assessment of the
103
Disposal Concept, NECSA Report GEA-1353 (NWS-RPT-00\013), South
African Nuclear Energy Corporation, Pretoria (2000).
104
APPENDIX C: ASSESSMENT MODEL
C.1
DESIGN SCENARIO
It is important to note that the release and migration processes described in Section
C.1.1 are assumed to occur only once the capsule containing the source container has
started to fail (see Section 4). Details concerning the mathematical model used to
represent the failure of capsule and other near-field engineered barriers are given in
Appendix C.3.
The processes identified in leading and off diagonal elements in the Interaction
Matrices in Figures 7 and 8 are listed in Table C.1. The associated equations are
listed in Table C.1 and discussed below.
C.1.1
Release Processes
As discussed in Section 5.2.1, the radionuclides in the source container could be in a
number of different physical and chemical forms and release of radionuclides could
occur on breaching of the waste capsule due to the following mechanisms.
Instantaneous dissolution of radionuclides that are in a form that would result in
immediate release to water once the capsule containing the source has failed (e.g.
liquid, soluble solid, surface contamination) (H-3, Ni-63, Sr-90, Cs-137, Pb-210, Ra226 and Am-241).
105
TABLE C.1. RELEASE AND MIGRATION PROCESSES AND ASSOCIATED EQUATIONS FOR THE DESIGN SCENARIO
System
Component
Near Field
Process
Equation/Comment
Dissolution
Once water contacts the source, it is assumed that radionuclides in the
source can be dissolved and transferred from the capsule containing the
source into the surrounding containment barrier/disposal zone due to
advection, dispersion and/or diffusion. Instantaneous dissolution congruent
release models are considered depending on the chemical and physical form
of the source.
Sorption
Equation 6, the following general formula:
Equation 8 to Equation 10,
Decay
Degradation
Advection
Dispersion
Equation 12 to Equation 15, Equation 17 and Equation 18
Equation 1
See Appendix K.3
Equation 6, Equation 10, Equation 14, and Equation 15
Implicitly represented through the discretisation of the geosphere into a
series of compartments and allowing compartment widths to increase
perpendicular to groundwater flow
Diffusion
Groundwater flow
Percolation
Geosphere
Sorption
Equation 12, Equation 13, Equation 17 and Equation 18
Equation 11 and Equation 16
Set equal to the minimum of the infiltration rate of water through the
unsaturated zone, and the hydraulic conductivity of the disposal zone and
containment barrier
the following general formula:
Equation 8, Equation 14, Equation 15, Equation 17, Equation 18 and
Equation 20
106
System
Component
Process
Equation/Comment
Decay
Advection
Dispersion
Equation 1
Equation 14 and Equation 15
Implicitly represented through the discretisation of the geosphere into a
series of compartments and allowing compartment widths to increase
perpendicular to groundwater flow
Equation 17, Equation 18, Equation 19 and Equation 20
Equation 16
Rate specified in Section 3.2.1
Not explicitly represented. Implicitly represented via percolation and
groundwater transport.
Equation 21
Not explicitly represented. Implicitly represented by modelling the
abstraction of water from the geosphere from consumption by humans and
assuming it is directed to the “elsewhere” compartment (Equation 21).
Not explicitly represented. Implicitly represented by not modelling the loss
of activity from the soil due to uptake by flora.
Not explicitly represented. Implicitly represented by assuming uniform
concentration of radionuclides in the soil.
Not explicitly represented. Implicitly represented by assuming uniform
concentration of radionuclides in the soil.
Diffusion
Groundwater flow
Percolation
Recharge
Biosphere
Abstraction
Excretion by humans
Excretion by animals
Bioturbation
Ploughing
Decay
Precipitation
Sorption
Suspension
Deposition onto soil
Deposition onto flora
Equation 1
Not explicitly represented. Implicitly represented via percolation in
unsaturated zone and groundwater flow in saturated zone.
Equation 24Equation 35
Equation 24
Not explicitly represented. Implicitly represented by not modelling the loss
of activity from the soil due to suspension.
Equation 32
107
System
Component

Process
Equation/Comment
Translocation
Root uptake
Erosion
Percolation
Death and decay of
crops
Cultivation
Harvesting
Rearing
Food preparation
losses
Equation 32
Equation 32
Equation 23
Equation 14
Not explicitly represented. Implicitly represented by not modelling the loss
of activity from the soil due to uptake by crops.
Considered by modelling the ingestion of crops by humans (Equation 31)
Considered by modelling the ingestion of crops by humans (Equation 31)
Considered by modelling the ingestion of animals by humans (Equation 33)
Equation 32
Congruent release of radionuclides that are in a form that would result in slow release to water (e.g. solid with low solublity) (Co-60).
It is recognised that the instantaneous dissolution and congruent release mechanisms could, under certain circumstances, be solubility limited
(see Table 15 and Appendix J of Generic Safety Assessment(GSA)). However, no solubility limitation is considered for the reference case
calculations (a conservative assumption).
For the congruent release model, it is assumed that once the engineered barriers containing the sealed source have failed, the source will begin to
corrode/dissolve and radionuclides become available for release. The fraction of the sealed source inventory released in any time period is equal
to the amount which becomes available divided by the inventory remaining in the source. For short time periods this simplifies to the rate of
change of availability with time divided by the amount which remains unavailable.
It is assumed that the source is a sphere of material whose radius decreases with time as it dissolves / corrodes. Ignoring decay (since this will
automatically be calculated by AMBER), the amount available at a time t, is therefore equal to:
108
Equation 2
4 3 4
3 
 r   (r  C r t ) 
3

I  3
4 3


r


3


where I is the radionuclide inventory (Bq), r is the initial radius of the source (m), Cr
is the corrosion / dissolution rate (m y-1).
Cancelling terms and expanding gives:
Equation 3
 r 3  r 3 3r 2 C r t 3rC r2 t 2 C r3t 3  
I   3   3 

 3  
3
3
r
r
r
r
r 


Therefore the amount available at time t equals:
Equation 4
 3r 2 C r t 3rC r2 t 2 C r3t 3 
I  

 3 
3
3
r
r
r 

Differentiating with respect to time gives the rate of change of the amount available
with time:
Equation 5
 3r 2 C r 6rC r2 t
3C r3t 2 

I   3 

r3
r 3 
 r
C.1.2 Liquid Migration Processes
Advective and Dispersive Release – Disposal in the Unsaturated Zone
The advective and dispersive transfer rate of radionuclides released from the capsule
containing the source due to percolation of water through the near field (UnsatLeach, in
y-1)is given by:
Equation 6
109
UnsatLeach 
q PERC f
Lc  wc Rc
where qPERC is the annual percolation rate through the capsule compartment (i.e. the
compartment representing the capsule that contains the source) (m y-1) (equal to the
minimum of the infiltration rate of water through the unsaturated zone, and the
hydraulic conductivity of the disposal zone and containment barrier), f is the fraction
of the waste that is available for release (unitless) (given by Equation 37, Equation
38 and Equation 39), Lc is the length of the capsule compartment in the direction of
water flow (m), wc is the water-filled porosity of the capsule compartment
(unitless), and Rc is the element dependent retardation of the capsule compartment
(unitless).
wc is calculated using the following general formula:
Equation 7
w   
where ε is the degree of saturation (unitless) in the compartment and θ is the total
porosity of the compartment (unitless). For the purposes of the GSA, it is assumed
that all the water-filled porosity contributes to flow and so total porosity and
effective porosity have the same values.
Rc is calculated using the following general formula:
Equation 8
R 1 
 Kd
w
where  is the dry bulk density of the compartment (kg m-3) and Kd is the sorption
coefficient of the element in the compartment (m3 kg–1).
 is calculated using the following general formula:
Equation 9
   g 1  
where gis the grain density of the compartment (kg m-3).
Advective and Dispersive Release - Disposal in the Saturated Zone
The advective transfer rate from the release of radionuclides from the capsule
containing the source due to the flow of groundwater through the near field (SatLeach,,
in y-1) is given by:
110
Equation 10
 SatLeach 
qc f
wc Lc Rc
where qc is the Darcy velocity of the groundwater through the capsule compartment
(m y-1), wc is the water-filled porosity of the capsule compartment (unitless), Lc is
the length of capsule compartment in the direction of groundwater flow (m), and
Rc(unitless) is the retardation factor in the capsule compartment for the radionuclide.
qc is calculated by:
Equation 11
qc   K c
H
x
where Kc (m y-1) is the hydraulic conductivity of the capsule compartment and H/x
is the hydraulic gradient (unitless).
The dispersion of radionuclides in the direction of groundwater movement
(longitudinal dispersion) is not represented explicitly as a mathematical model. This
is because when a flow path is divided into a number of equally sized compartments
in the direction of groundwater flow, the mathematical representation as a series of
well-mixed compartments introduces dispersion. The effective Peclet number (a
measure of dispersion) is twice the number of compartments in the flow path (see
discussion in Appendix B of [K.1]). Where the compartments are not of the same
size, the effective Peclet number is dominated by the largest compartment.
Diffusive Release – Disposal in the Unsaturated and Saturated Zones
The diffusive release from the capsule compartment (DiffRelF,, in y-1) is given by:
Equation 12
 DiffRelF

Adiff f DEffc
Rc Vc  c  wc
where Adiff (m2) is the cross-sectional area relevant to the diffusive release from the
capsule, f is the fraction of the waste that is available for release, DEffc (m2 y-1) is the
effective diffusion coefficient for the capsule compartment, Rc(unitless) is the
111
retardation factor in the capsule compartment for the radionuclide, Vc (m3) is the
volume of the capsule compartment, c (m) is a representative diffusion length
between the capsule compartment and the adjacent compartment, generally taken to
be the distance between the mid-points of the compartments in the direction of the
diffusive flux, and wc is the water-filled porosity of the capsule compartment.
In addition to this ‘forward’ diffusive transfer rate, there is a need to represent a
corresponding ‘backward’ diffusive transfer rate in the reverse direction from the
compartment adjacent to the capsule compartment (DiffRelB,, in y-1). This transfer rate
is given by:
Equation 13
 DiffRelB

Adiff f DEffA
R A V A  c  wA
where DEffA (m2 y-1) is the effective diffusion coefficient for the compartment
adjacent to the capsule compartment, RA(unitless) is the retardation factor for the
radionuclide in the compartment adjacent to the capsule compartment, VA (m3) is the
volume of the compartment adjacent to the capsule compartment, c (m) is a
representative diffusion length between the adjacent compartment and the capsule
compartment, generally taken to be the distance between the mid-points of the
compartments in the direction of the diffusive flux, and wA is the water-filled
porosity of the compartment adjacent to the capsule compartment.
Advective and Dispersive Transport – Disposal in the Unsaturated Zone
The transfer rate of contaminated water percolating (due to advection and dispersion)
through the unsaturated near field, unsaturated geosphere and soil (PERC, in y-1) is
given by:
Equation 14
 PERC 
q PERC
Lw R
where qPERC is the annual percolation rate through the compartment (m y-1), L is the
length of the compartment in the direction of water flow (m), w is the water-filled
porosity of the compartment (unitless), and R is the element dependent retardation of
the compartment (unitless) (given by Equation 7). For flow in a fracture it is
assumed that there is no retardation, and so R is unity
112
Advective and Dispersive Transport in the Saturated Zone
For transport through the saturated zone, the advective transfer rate (A, in y-1)is
given by:
Equation 15
A 
q
w L R
where q is the Darcy velocity of the groundwater in the compartment (m y-1), w is
the water-filled porosity of the compartment (unitless), L is the length of the
compartment in the direction of water flow, and R is the element dependent
retardation of the compartment (unitless).
q is given by:
Equation 16
q  K
H
x
where K (m y-1) is the hydraulic conductivity of the compartment and H/x is the
hydraulic gradient (unitless).
As discussed above, the dispersion of radionuclides in the direction of groundwater
movement (longitudinal dispersion) is implicitly represented through the
discretisation of the saturated zone into a series of compartments. Contaminant
dispersion at right angles to the direction of groundwater movement in the saturated
medium (transverse dispersion) is not represented explicitly as a process because the
compartment dimensions can be defined to represent the increase in plume
dimensions due to lateral spreading.
Diffusive Transport – Disposal in the Unsaturated and Saturated Zones
The ‘forward’ diffusive transfer rate (DiffD, in y-1) is given by:
Equation 17
 DiffD

AU D EffU
RU VU  U  wU
where AU (m2) is the cross-sectional area relevant to the diffusive transfer from the
upstream compartment, DEffU (m2 y-1) is the effective diffusion coefficient for the
upstream compartment, RU(unitless) is the retardation factor in the upstream
compartment for the radionuclide, VU (m3) is the volume of the upstream
compartment, U (m) is a representative diffusion length between the upstream and
113
downstream compartments, generally taken to be the distance between the midpoints of the compartments in the direction of the diffusive flux, and wU is the
water-filled porosity of the upstream compartment.
In addition to this ‘forward’ diffusive transfer rate, there is a need to represent a
corresponding ‘backward’ diffusive transfer rate in the reverse direction (DiffU,, in
y-1). This transfer rate is given by:
Equation 18
DiffU

AU DEffD
RD VD U wD
where AU (m2) is the cross-sectional area relevant to the transport, DEffD (m2 y-1) is
the effective diffusion coefficient for the downstream compartment, RD(unitless) is
the retardation factor for the radionuclide in the downstream compartment, VD (m3)
is the volume of the downstream compartment, U (m) is a representative diffusion
length between the upstream and downstream compartments, generally taken to be
the distance between the mid-points of the compartments in the direction of the
diffusive flux, and wD is the water-filled porosity of the downstream compartment.
The diffusive transfer rate from a fractured compartment into a matrix compartment
(rm, in y-1) is given by:
Equation 19
 rm 
2 a DEffm
 wf 
where a is the flow wetted surface area per unit volume of rock (m2 m-3),  wfis the
water-filled fracture porosity (unitless), DEffmis the effective diffusion coefficient of
the matrix compartment (m2 y-1), and  is the depth of the matrix compartment (m).
The reverse transfer rate from a matrix compartment back to the fracture (mr, in y-1)
is given by:
Equation 20
 mr 
2 DEffm
Rm  2 wm
where Rm is the retardation coefficient of the radionuclide in the matrix, and wm is
the water-filled matrix porosity (unitless).
114
Water Abstraction
The transfer rate of radionuclides in groundwater abstracted from the geosphere to
soil due to irrigation of crops (irrig, in y-1) is given by:
Equation 21
irrig 
Dil Virrig
 ww Vw Rw
where Dil is the fraction of the water demand supplied by contaminated water, Virrig
is the volume of irrigation water that reaches the soil (m3 y-1), ww is the water-filled
porosity of the compartment from which the water is abstracted (unitless), Vw is the
volume of the compartment from which the water is abstracted (m3), Rw is the
retardation coefficient (unitless) of the compartment from which the water is
abstracted.
For this abstraction, it is necessary to consider only the volume of irrigation water
reaching the soil since it represents the transfer of radionuclides from the geosphere
to the soil (rather than to the crops). Some of the water abstracted from the
geosphere for irrigation purposes will be intercepted by the crops and will not reach
the soil since it is either taken up directly into the crop or evaporated from the crop
surface. This water is accounted for in the other water abstraction considered below.
The transfer rate of radionuclides due to abstraction of water for watering of animals
and domestic purposes (other, in y-1) is given by:
Equation 22
other 
Dil Vother
ww Vw Rw
where Dil is the fraction of the water demand supplied by contaminated water, Vother
is the volume of water abstracted for watering of animals and domestic purposes
(includes the volume of irrigation water not reaching the soil due to interception by
crops) (m3 y-1), ww is the water filled porosity of the compartment from which the
water is abstracted (unitless), Vw is the volume of the compartment from which the
water is abstracted (m3), and Rw is the retardation coefficient (unitless) of the
compartment from which the water is abstracted.
Erosion
The transfer rate of radionuclides by erosion of a compartment (λEROS, in y-1) is given
by:
115
Equation 23
 EROS 
d EROS
D
where dEROS is the erosion rate for the compartment (m y-1), and D is the depth of the
compartment from which erosion takes place (m).
Suspension
The suspension of dust above a soil compartment is modelled by using dust loading
factors, where the concentration of a radionuclide in the air above the soil, CAir (Bq
m-3) is given by:
Equation 24
C Air   Dry
( RSoil  1)
c Dust
RSoil
where χDryis the radionuclide concentration in the dry surface soil (Bq kg-1 dry
weight soil), RSoil is the retardation coefficient for soil compartment (unitless), cDustis
the dust level in the air above the soil compartment (kg m-3).
χDry is given by:
Equation 25
 Dry 
C Soil
 Soil
where CSoil is the radionuclide concentration in the soil (Bq m-3) and Soil is the dry
bulk density of the soil (kg m-3).
CSoil is given by:
Equation 26
C Soil 
Amount Soil
VSoil
where AmountSoil is the amount of the radionuclide in the soil (Bq) and VSoil is the
volume of the compartment representing the soil (m3).
C.1.3 Exposure Mechanisms
116
For the Design Scenario, it is assumed that exposure can only occur once the capsule
has started to fail and the institution control period has ended (Section 4.2). Details
concerning the mathematical model used to represent the failure of capsule and other
near-field engineered barriers are given in Appendix C.3 and it is assumed that the
institutional control period ends 50 years after site closure (Section 4.2).
The exposure mechanisms identified in the off diagonal elements in the Interaction
Matrices in Figures 7 and 8 are listed in Table C.2. Equations are given below that
are used to calculate the annual effective dose received by an average adult member
of an exposure group from these exposure mechanisms.
TABLE C.2. EXPOSURE MECHANISMS AND ASSOCIATED EQUATIONS
FOR THE DESIGN SCENARIO
Mechanism
Ingestion
Equation
Inhalation
Medium
Groundwater
Soil
Crops
Animals
Dust
External Irradiation
Soil
Equation 36
Equation 27
Equation 29
Equation 31
Equation 33
Equation 35
Ingestion of Groundwater
The annual individual effective dose to a human from the consumption of drinking
water (DWat, in Sv y-1) is given by:
Equation 27
DW at  CW Ing W at DC Ing
where CW is the radionuclide concentration in the abstracted water (Bq m-3), IngWat is
the individual ingestion rate of water (m3 y-1), and DCIng is the dose coefficient for
ingestion (Sv Bq-1).
CW is given by:
Equation 28
CW 
Dil Amountw
 w Vw Rw
where Dil is the contribution of the water from the abstraction borehole to the total
water ingested, Amountw is the amount of the radionuclide in the compartment from
which the water is abstracted (Bq), w is the water-filled porosity of the
117
compartment (unitless), Vw is the volume of the compartment from which the water
is abstracted (m3), and Rw is the retardation coefficient of the compartment from
which the water is abstracted (unitless).
Ingestion of Soil
Soil can be inadvertently ingested by humans. The annual individual dose to a
human from the ingestion of soil (DSed, in Sv y-1) is given by:
Equation 29
D Sed   W et Ing Sed OOut DC Ing
where χWet is the radionuclide concentration in the soil (Bq kg-1 wet weight), IngSed is
the individual inadvertent ingestion rate of soil (kg wet weight h-1), OOut is the
individual occupancy on the soil (h y-1), and DCIng is the dose coefficient for
ingestion (Sv Bq-1).
χWet is given by:
Equation 30
 Wet 
C Soil
 Soil  wSoil Wat
where Csoil is the radionuclide concentration in the soil (Bq m-3), Soil is the dry bulk
density of the soil (kg m-3), wSoil is the water filled porosity of the soil (unitless),
and Watthe density of water (kg m-3).
Ingestion of Crops
The annual individual effective dose to a human from the consumption of a crop,
(DCrop, in Sv y-1), is given by:
Equation 31
DCrop   Crop Ing Crop DC Ing
where  Crop is the radionuclide concentration in the crop (Bq kg-1 fresh weight of
crop), IngCropis the individual ingestion rate of the crop (kg fresh weight y-1), and
DCIng is the dose coefficient for ingestion (Sv Bq-1).
The χcrop term is calculated using:
Equation 32
118
 Crop  (CFCrop  (1  f Pr ep ) sCrop )  Dry   Crop d Irr CW
(1  f Pr ep )(1  f Trans ) e
T Wcrop
 f Trans
YCrop
where CFcrop is the concentration factor for the crop (Bq kg-1 fresh weight of crop/Bq
kg-1 (dry weight of soil)), fPrep is the fraction of external contamination on the crop
lost due to food processing (unitless), sCrop is the soil contamination on the crop (kg
dry weight soil kg-1 fresh weight of crop), χDryis the radionuclide concentration in the
dry surface soil (Bq kg-1 dry weight soil), cropis the interception fraction for
irrigation water on the crop (unitless), dIrr is the depth of irrigation water applied to
the crop (m y-1), CWis the radionuclide concentration in the abstracted water (Bq m3
), fTrans is the fraction of activity transferred from external to internal plant surfaces
(unitless), T is the interval between irrigation and harvest (y), Wcrop is the removal
rate of irrigation water from the crop by weathering processes (weathering rate) (y-1),
and YCrop is the yield of the crop (kg fresh weight of crop m-2 y-1).
Ingestion of Animals
The annual individual effective dose to a human from the consumption of animal
produce (DAnm, in Sv y-1) is given by:
Equation 33
D Anm   Anm Ing Anm DC Ing
where χAnm is the radionuclide concentration in the animal product (Bq kg-1 fresh
weight of product), IngAnm is the individual consumption rate of the animal product
(kg fresh weight of product y-1) and DCIng is the dose coefficient for ingestion
(Sv Bq-1).
The χAnm term is calculated using:
Equation 34
 Anm  CFAnmCW Ing AW
where CFAnmis the concentration factor for the animal product (d kg-1 fresh weight of
product), Cw is the radionuclide concentration in the water used for watering animals
(Bq m-3) and IngAW is the consumption rate of water by the animal (m3 d-1).
Inhalation of Dust
The annual individual dose to a human from the inhalation of dust (DDust, in Sv y-1)
is given by:
119
Equation 35
DDust  C Air OOut InhSed DC Inh
whereCAir is the radionuclide concentration in the air above the soil (Bq m-3), InhSed
is the breathing rate of the human on the contaminated soil (m3 h-1), and DCInh is the
dose coefficient for inhalation (Sv Bq-1).
External Irradiation
The annual individual dose to a human from external irradiation from soil (DExSoil, in
Sv y-1) is given by:
Equation 36
DExSoil  C Soil OOut DC Exts
where CSoil is the concentration in the soil (Bq m-3), OOut is the individual occupancy
outdoors on the contaminated soil (h y-1), and DCExtsis the dose coefficient for
external irradiation from soil (Sv h-1/Bq m-3).
C.2
DEFECT SCENARIO
The mathematical model for this scenario is the same as that for the Design Scenario
discussed in Appendix C.1, although some different parameter values are used
(Appendix D).
C.3
REPRESENTING NEAR-FIELD DEGRADATION
It is necessary to consider the degradation of the following near-field components
(see Table 11):
the stainless steel capsule that contains the source container (the source container is
assumed to have failed before disposal) and the stainless steel disposal
container that contains the capsule; and
the containment barrier, the disposal zone backfill and plug, the closure zone
backfill, and the disturbed zone backfill.
Degradation can affect both the physical and chemical performance of the near-field
components.
C.3.1 Physical Performance
Capsule and Disposal Container
Failure times for each component are specified in Appendix D based upon the
corrosion modelling reported in Appendix I. The physical performance of each of
120
these components could fail in a linear manner over a period of time. It could start at
a user-defined time (tPhysDegStart, in y) (when the water/gas tightness of the component
is first breached) and end at a user-defined time (tPhysDegEnd, in y) (when the
component has totally failed and is fully degraded). Between these two times, linear
failure could be assumed. However, corrosion model results discussed in Appendix I
indicate that the physical performance of each components can be consider to occur
essentially instantaneously. Nevertheless, flexibility in the model is maintained by
adopting the linear failure model but setting tPhysDegEnd to be marginally greater than
tPhysDegStart.
Prior to the start of the failure of the stainless steel capsule none of the waste is
available for release. However, once the capsule starts to fail (at time tCapPhysDegStart,
in y), the fraction of the waste available for release is assumed to start to increase in
a linear manner until all of the waste is assumed to be available once the capsule is
fully degraded ((at time tCapPhysDegEnd, in y). Thus, the value of the fraction of waste
available for release (f, unitless) (as used in Equation 6, Equation 10,
Equation 12 and Equation 13) is a function of time:
Equation 37
f t   0
t < tCapPhysDegStart
f t   1
t ≥ tCapPhysDegEnd
Equation 38
Equation 39
f t  
t  t CapPhysDegStart
t CapPhysDegEnd  t CapPhysDegStart
otherwise
Containment Barrier and Backfill Material
The hydraulic conductivity and total porosity of the cement grout containment
barrier and backfill material are assumed to increase due to physical degradation
(e.g. cracking) and chemical degradation (e.g. calcium leaching and sulphate attack)
for all scenarios (see Appendix H). These changes can be represented by the
definition of ‘undegraded’ and ‘degraded’ values for both parameters. A function,
fPhysDeg, can be used to describe the transition of the values from the undegraded state
prior to the start of degradation (at time tMatPhysDegStart , in y) to the end of degradation
(at time tMatPhysDegEnd , in y):
Equation 40
121
f PhysDeg (t )  0
t < tMatPhysDegStart
Equation 41
f PhysDeg (t )  1 t ≥ tMatPhysDegEnd
Equation 42
f PhysDeg (t ) 
t  t MatPhysDegStart
t MatPhysDegEnd  t MatPhysDegStart
otherwise
The value ofhydraulic conductivity at a given time (K(t), in m y-1) can be determined
using the function as follows:
Equation 43
K (t )  (1  f PhysDeg (t )) K UnDeg  f PhysDeg (t ) K Deg
where KUnDeg and KDeg are the undegraded and degraded hydraulic conductivities,
respectively (both in m y-1).
The same approach can be used to calculate the total porosity (θ, unitless):
Equation 44
 (t )  (1  f PhysDeg (t ))  UnDeg  f PhysDeg (t )  Deg
whereθ UnDeg and θ Deg are the undegraded and degraded total porosities, respectively
(both unitless).
C.4.2 Chemical Performance
Degradation of the chemical performance of cement grout containment barrier and
backfill material is assumed to occur in all scenarios. The processes that lead to
chemical degradation, such as calcium leaching and sulphate attack, are implicitly
rather than explicitly modelled. Failure times for each component are specified in
Appendix D based upon cement grout degradation model presented in Apppendix H.
It is assumed that the chemical performance of each of these components does not
degrade instantaneously; degradation is assumed to occur in a linear manner over a
period of time. It is assumed to start at a user-defined time (tChemDegStart, in y) and end
at a user-defined time (tChemDegEnd, in y). Between these two times, linear degradation
is assumed.
It is assumed that chemical evolution of the cement grout near field affects the nearfield sorption coefficients. An approach similar to that used for modelling the change
in hydraulic conductivity and porosity is used to represent changes in sorption
coefficients. A function, fChemDeg, can be used to describe the transition of the values
122
from the undegraded state prior to the start of degradation (at time tMatChemDegStart, in
y) to the end of degradation (at time tMatChemDegEnd, in y):
Equation 45
f ChemDeg (t )  0
t < tMatChemDegStart
f ChemDeg (t )  1
t ≥ tMatChemDegEnd
Equation 46
Equation 47
f ChemDeg (t ) 
t  t MatChemDegStart
t MatChemDegEnd  t MatChemDegStart
otherwise
The value of a radionuclide’snear-field sorption coefficient at a given time (Kd(t), in
m3 kg-1) can be determined using the function as follows:
Equation 48
K d (t )  (1  f ChemDeg (t )) K dUnDeg  f ChemDeg (t ) K dDeg
where KdUnDeg and KdDeg are the undegraded and degraded sorption coefficients,
respectively (both in m3 kg-1).
References for Appendix C
C.1 PENFOLD, J.S.S., LITTLE, R.H., ROBINSON, P.C., AND SAVAGE, D.,
Improved Safety Assessment Modelling of Immobilised LLW Packages for
Disposal, Ontario Power Generation Technical Report 05386-REP-03469.310002-R00, Toronto (2002).
123
APPENDIX D: ASSESSMENT DATA
Table D.1 lists the parameters used in the mathematical models described in
Appendix C and identifies the table in which the associated data can be found.
TABLE D.1. PARAMETERS FOR THE MATHEMATICAL MODEL AND
LOCATION OF ASSOCIATED VALUES
Symbol
Δc
ΔU
δ
H/x

θ
θDeg
θH
θUndeg
w
wA
wB
wc
wDC
wf
wm
wSoil
wU
ww
λA
λDiffD
λDiffRelB
λDiffRelF
λDiffU
EROS
λirrig
mr
λN
other
Perc
rm
Definition
a representative diffusion length between the
upstream and downstream compartments
a representative diffusion length between the
capsule compartment and the adjacent
compartment
depth of the matrix compartment
hydraulic gradient
degree of saturation
total porosity
total porosity of degraded cement grout
total porosity of the base of the house
total porosity of undegraded cement grout
water-filled porosity
water-filled porosity of the compartment
adjacent to the capsule compartment
water-filled porosity of the downstream
compartment
water-filled porosity of capsule compartment
water-filled porosity of the contaminated drill
core
water-filled fracture porosity
water-filled matrix porosity
water-filled porosity of the soil
water-filled porosity of the upstream
compartment
water-filled porosity of the compartment
from which the water is abstracted
advective transfer rate in the saturated zone
forward (downstream) diffusive transfer rate
backward diffusive transfer rate from the
compartment adjacent to the capsule
forward diffusive release rate from capsule
compartment
Units
m
backward (upstream) diffusive transfer rate
transfer rate due to erosion
transfer rate in groundwater abstracted from
the geosphere to soil due to irrigation of crops
diffusive transfer rate from matrix to fracture
decay constant for radionuclide N
transfer rate due to abstraction of water for
watering of animals and domestic purposes
advective and dispersive transfer rate through
the unsaturated zone
diffusive transfer rate from fracture to matrix
Value
Tables D.15, D.19 and D.20
m
Table D.15
m
-
Tables D.19 and D.20
Tables D.13 and D.18
Tables D.13, D.18, D.22
Tables D.13, D.18, D.22
Table D.13
Table D.22
Table D.13
Calculated using Equation 10
Calculated using Equation 10
-
Calculated using Equation 10
-
Calculated using Equation 10
Calculated using Equation 10
-
Calculated using Equation 10
Calculated using Equation 10
Calculated using Equation 10
Calculated using Equation 10
-
Calculated using Equation 10
y-1
y-1
y-1
Calculated using Equation 15
Calculated using Equation 17
Calculated using Equation 13
y-1
Calculated using
y-1
y-1
y-1
Equation 12
Calculated using Equation 18
Calculated using Equation 23
Calculated using Equation 21
y-1
y-1
y-1
Calculated using Equation 20
Table D.4
Calculated using Equation 21
y-1
Calculated using Equation 14
y-1
Calculated using Equation 19
124
Symbol
λSatLeach
λUnsatLeach
v
crop

Bh
g
Soil
Wat
ΧAnm
Definition
advective and dispersive transfer rate from
capsule containing the source due to
groundwater flow
advective and dispersive transfer rate from
capsule containing the source due to water
percolation
ventilation rate of the house
interception fraction for irrigation water on
the crop
dry bulk density
dry bulk density of the borehole’s disposal
zone
grain density
dry bulk density of the soil
density of water
radionuclide concentration in the animal
product
χCrop
radionuclide concentration in the crop
χDry
radionuclide concentration in the surface soil
χWet
radionuclide concentration in the surface soil
A
CFAnm
flow wetted surface area per unit volume of
rock
cross-sectional area of the disposal borehole
cross-sectional area relevant to the diffusive
release from the capsule
amount of a radionuclide in a compartment
amount of Ra-226 in the borehole’s disposal
zone
amount of a radionuclide in the soil
amount of a radionuclide in the compartment
from which the water is abstracted
cross-sectional area relevant to the diffusive
transfer from the upstream compartment
breathing rate of the human in the house
concentration of a radionuclide in the air
above the soil
dust level in the air above the soil
compartment
concentration factor for the animal product
CFcrop
concentration factor for the crop
Cr
CSoil
CW
corrosion/dissolution rate of source
radionuclide concentration in the soil
radionuclide concentration in the abstracted
Ab
Adiff
Amount
AmountRaBh
AmountSoil
Amountw
AU
BRgas
CAir
cDust
Units
y-1
Value
Calculated using Equation 10
y-1
Calculated using Equation 6
y-1
-
Table D.22
Table D.24
kg m-3
kg m-3
Calculated using Equation 9
Calculated using Equation 9
kg m-3
Tables D.13, D.18,D.22, D.5
and D.7
Calculated using Equation 9
1000 kg m-3
Calculated using Equation 34
kg m-3
kg m-3
Bq kg-1 fresh
weight of
product
Bq kg-1 fresh
weight
Bq kg-1 dry
weight
Bq kg-1 wet
weight
m2 m-3
m2
m2
Calculated using Equation 32
Calculated using Equation 25
Calculated using Equation 30
Tables D.2 and D.3
Table D.14
Table D.15
Bq or moles
Bq
Calculated using Equation 1
Calculated using Equation 1
Bq
Bq
Calculated using Equation 1
Calculated using Equation 1
m2
Tables D.2 and D.3
m3 h-1
Bq m-3
Table D.23
Calculated using Equation 24
kg m-3
Table D.22
d kg-1 fresh
weight of
product
Bq kg-1 fresh
weight of
crop/Bq kg-1
(dry weight of
soil)
m y-1
Bq m-3
Bq m-3
Table D.12
Table D.8
Table D.13
Calculated using Equation 26
Calculated using Equation 28
125
Symbol
fChemDeg
Definition
water
depth of the compartment from which erosion
takes place
annual individual effective dose to a human
from the consumption of animal produce
thickness of the borehole’s closure zone
dose coefficient for external irradiation from
soil
dose coefficient for ingestion
dose coefficient for inhalation
annual individual effective dose to a human
from the consumption of a crop
annual individual dose to a human from the
inhalation of dust
effective diffusion coefficient for the
compartment adjacent to the capsule
compartment
effective diffusion coefficient for the capsule
compartment
effective diffusion coefficient for the
downstream compartment
effective diffusion coefficient for the matrix
compartment
effective diffusion coefficient for the
upstream compartment
erosion rate for the compartment
annual individual dose to a human from
external irradiation from soil
contribution of water from abstraction
borehole to the total water demand
depth of irrigation water applied to the crop
annual individual dose to a human from the
ingestion of soil
annual individual effective dose from the
consumption of drinking water
fraction of the waste that is available for
release
extent of chemical degradation
fPhysDeg
extent of physical degradation
-
fPrep
fraction of external contamination on the crop
lost due to food processing
fraction of activity transferred from external
to internal plant surfaces
disposed inventory of the radionuclide,
decay-corrected to the start time of the
capsule’s physical failure
-
Calculated using Equation 37 to
Equation 39
Calculated using Equation 45 to
Equation 47
Calculated using Equation 40 to
Equation 42
Table D.11
-
Table D.10
D
DAnm
dBh
DCExts
DCIng
DCInh
DCrop
DDust
DEffA
DEffc
DEffD
DEffm
DEffU
dEROS
DExSoil
Dil
dIrr
DSed
DWat
f
fTrans
Ig
IngAnm
individual ingestion rate of animal product
Units
m
Sv y-1
m
Sv h-1/Bq m-3
Value
Table D.22
Calculated using Equation 33
Table D.14
Table D.5
Sv Bq-1
Sv Bq-1
Sv y-1
Table D.5
Table D.5
Calculated using Equation 31
Sv y-1
Calculated using Equation 35
m2 y-1
Table D.16
m2 y-1
Table D.16
m2 y-1
Table D.16
m2 y-1
Table D.16
m2 y-1
Table D.16
m y-1
Sv y-1
Table D.22
Calculated using Equation 36
-
Table D.21
m y-1
Sv y-1
Table D.24
Calculated using Equation 29
Sv y-1
Calculated using Equation 27
-
Bq
kg fresh
weight of
product y-1
Calculated from I e- λt where I is
the initial inventory disposed
(1E+12 Bq, Section 3.1.1), λ (y1
) is the decay constant (Table
L.4), and t (y) is the start time of
the capsule’s physical failure
(Tables D.16 and D.17)
Table D.23
126
Symbol
IngAW
IngCrop
Definition
consumption rate of water by the animal
individual ingestion rate of the crop
IngSed
individual inadvertent ingestion rate of soil
IngWat
InhSed
individual ingestion rate of freshwater
breathing rate of the human on the
contaminated soil
hydraulic conductivity of a medium
K
Kc
Kd
KdDeg
KDeg
KdUnDeg
KUnDeg
L
Lc
Ogas
OOut
Q
qc
qPERC
r
R
RA
Rc
RD
Rm
RSoil
RU
Rw
sCrop
hydraulic conductivity of the capsule
compartment
sorption coefficient of the element in the
compartment
sorption coefficient of degraded cement grout
Hydraulic conductivity of degraded cement
grout backfill
sorption coefficient of undegraded cement
Hydraulic conductivity of undegraded cement
grout backfill
length of compartment in the direction of
water flow
length of capsule compartment in the
direction of water flow
individual occupancy in the house
The individual occupancy on the soil
Darcy velocity of groundwater through a
compartment
Darcy velocity of groundwater through the
capsule compartment
annual percolation rate through the capsule
compartment
initial radius of the source
element dependent retardation of the
compartment
element dependent retardation of the
compartment adjacent to the capsule
compartment
element dependent retardation of the capsule
compartment
element dependent retardation of the
downstream compartment
element dependent retardation of the matrix
compartment
element dependent retardation of the soil
compartment
element dependent retardation of the
upstream compartment
element dependent retardation of the
compartment from which the domestic and
agricultural water is abstracted
soil contamination on the crop
Units
m3 d-1
kg fresh
weight y-1
kg wet weight
h-1
m3 y-1
m3 h-1
m y-1
m y-1
m3 kg –1
Value
Table D.23
Table D.23
Table D.23
Table D.23
Table D.23
Tables D.13 and D.18
For cement grout in near-field
calculated using Equation 43
Table D.13
m3 kg –1
m y-1
Table D.7
For cement grout in near-field
calculated using Equation 48
Table D.7
Table D.13
m3 kg –1
m y-1
Table D.7
Table D.13
m
Tables D.15, D.19 and D.20
m
Table D.15
h y-1
h y-1
m y-1
Table D.23
Table D.23
Calculated using Equation 16
m y-1
Calculated using Equation 11
m y-1
Table D.13
m
-
Table D.13
Calculated using Equation 11
-
Calculated using Equation 11
-
Calculated using Equation 11
-
Calculated using Equation 11
-
Calculated using Equation 11
-
Calculated using Equation 11
-
Calculated using Equation 11
-
Calculated using Equation 11
kg dry weight
Table D.24
127
Symbol
T
tCapPhysDegStart
tCapPhysDegEnd
tChemDegStart
tChemDegEnd
tDelay
tMatChemDegStart
tMatChemDegEnd
tMatPhysDegStart
tMatPhysDegEnd
tPhysDegStart
tPhysDegEnd
V
Definition
is the interval between irrigation and harvest
time at which failure of the capsule’s physical
performance starts
time at which failure of the capsule’s physical
performance ends
time at which failure of a barrier’s chemical
performance starts
time at which failure of a barrier’s chemical
performance starts
average radon travel time from the soil into
the house
time at which failure of the cement grout
containment barrier’s and cement grout
backfill’s chemical performance starts
time at which failure of the cement grout
containment barrier’s and cement grout
backfill’s chemical performance ends
time at which failure of the containment
barrier and backfill material’s physical
performance starts
time at which failure of the containment
barrier and backfill material’s physical
performance ends
time at which failure of a barrier’s physical
performance starts
time at which failure of a barrier’s physical
performance ends
volume of the compartment
Units
soil kg-1 fresh
weight of crop
y
y
Table D.24
Tables D.16 and L.17
y
Tables D.16 and D.17
y
Tables D.16 and D.17
y
Tables D.16 and D.17
y
Table D.22
y
Tables D.16 and D.17
y
Tables D.16 and D.17
y
Tables D.16 and D.17
y
Tables D.16 and D.17
y
Tables D.16 and D.17
y
Tables D.16 and D.17
m3
m3 y-1
Derived from dimensions,
Tables D.15, D.19, D.20 and
D.22
Derived from dimensions, Table
D.15
Derived from dimensions,
Section 3.1.2
Derived from dimensions, Table
5
Derived from dimensions,
Tables D.15, D.19 and D.20
Table D.24
Derived from dimensions, Table
D.22
Table D.22
m3 y-1
Table D.22
m3
VBh
volume of the compartment adjacent to the
capsule compartment
volume of the borehole’s disposal zone
Vc
volume of the capsule compartment
m3
VD
volume of the downstream compartment
m3
VDC
Vh
volume of the contaminated drill core
total volume of the house
m3
m3
Virrig
is the volume of irrigation water that reaches
the soil
volume of water abstracted for watering of
animals and domestic purposes (includes the
volume of irrigation water not reaching the
soil due to interception by crops)
volume of the compartment representing the
soil
volume of compartment from which water is
abstracted
volume of the upstream compartment
VA
Vother
VSoil
Vw
VU
Value
m3
m3
m3
m3
Derived from dimensions, Table
D.22
Derived from dimensions,
Tables D.19 and D.20
Derived from dimensions,
Tables D.15, D.19 and D.20
128
Symbol
Wcrop
YCrop
D.1
Definition
removal rate of irrigation water from the crop
by weathering processes (weathering rate)
yield of the crop
Units
y-1
kg (fresh
weight of
crop) m-2 y-1
Value
Table D.9
Table D.24
INVENTORY AND RADIONUCLIDE DATA
TABLE D.2. RADIONUCLIDES DISPOSED AND ASSOCIATED DAUGHTERS
CONSIDERED
Disposed
Radionuclide
(1)
Co-60
Sr-90
I-129
Cs-137
Ra-226
Am-241
Cf-252
Short-lived
Daughter(s)
(2)
Daughter(s)
*
*
*
Pb-210*
Po-210
Np-237 Pa-233 U-233 Th-229*
→(branching ratio 0.9691)Cm-248→(branching ratio
0.9161)Pu-244*→Pu-240→U236→Th-232→Ra228*→Th-228*
Notes
1. For each disposed radionuclide, an inventory of 1 TBq per waste package is
assumed (see Section 3.1.1). It is also assumed that there are 50 waste packages
per borehole (see Section 3.1.2), giving a total inventory of 50 TBq for each
radionuclide.
2. * indicates a daughter with a half-life of less than 25 days (see Table D.3)
129
TABLE D.3. SHORT-LIVED DAUGHTERS WITH HALF-LIVES OF LESS
THAN 25 DAYS ASSUMED TO BE IN SECULAR EQUILIBRIUM WITH
THEIR PARENTS
Parent
Sr-90
Cs-137
Ra-226
Short Lived Daughters
 Y-90
 (branching ratio 0.94) Ba-137m
 Rn-222 Po-218 (branching ratio 0.9998) Pb-214  Bi-214  (branching ratio 0.9998) Po-214

(branching ratio 0.0002) At-218  Bi-214  (branching ratio 0.9998) Po-214
Ac-227
→(branching ratio 0.0138) Fr-223→( branching ratio 0.9862) Th-227→Ra-223→Rn-219→Po215→Pb-211→Bi-211
Th-229

Ra-225 Ac-225 Fr-221 At-217 Bi-213 (branching ratio 0.9784) Po-213 Pb-209

(branching ratio 0.0216) Tl-209 Pb-209
Ra-228
Th-228
Pu-244
→Ac-228
→Ra-224→Rn-220→Po-216→Pb-212→Bi-212→(branching ratio 0.641)Po-212
→ (branching ratio 0.359)Tl-208
→ (branching ratio 0.9988)U-240→Np-240m→(branching ratio 0.0011)Np-240
130
TABLE D.4. RADIONUCLIDE HALF-LIVES AND DECAY RATES
Radionuclide
Half-life
(y) (1)
5.27E+00
2.91E+01
1.57E+07
3.00E+01
2.23E+01
3.79E-01
5.75E+00
1.60E+03
5.75E+00
1.91E+00
7.34E+03
1.40E+10
7.39E-02
1.59E+05
2.34E+07
2.14E+06
4.32E+02
6.54E+03
8.26E+07
3.39E+05
2.64E+00
Co-60
Sr-90
I-129
Cs-137
Pb-210
Po-210
Ra-228
Ra-226
Ra-228
Th-228
Th-229
Th-232
Pa-233
U-233
U-236
Np-237
Am-241
Pu-240
Pu-244
Cm-248
Cf-252
Decay rate
(y-1) (2)
1.32E-01
2.38E-02
4.41E-08
2.31E-02
3.11E-02
1.83E+00
1.21E-01
4.33E-04
1.21E-01
3.63E-01
9.44E-05
4.95E-11
9.38E+00
4.36E-06
2.96E-08
3.24E-07
1.60E-03
1.06E-04
8.39E-09
2.04E-06
2.63E-01
Notes
1. Data from [1].
2. Decay constant =
ln 2
half life
References for Table D.4
[1]
ICRP (1983).
Radionuclide Transformations Energy and Intensity of
Emissions.International Commission on Radiological Protection, ICRP
Publication 38. Pergamon Press, Oxford.
131
TABLE D.5. RADIONUCLIDE DOSE COEFFICIENTS FOR INGESTION,
INHALATION AND EXTERNAL IRRADIATION
Radionuclide
Dose Coefficients for Adults (1)
Ingestion
Inhalation
External Irradiation
(Sv Bq-1)
(Sv Bq-1)
from soil
(2)
(2)
(Sv h-1/Bq m-3)
(3)
Water
Immersion
(Sv h-1/Bq m-3)
(4)
Co-60
3.4E-09
1.0E-08
3.0E-13
9.3E-13
Sr-90
3.1E-08
3.8E-08
7.9E-16
4.0E-15
I-129
1.1E-07
3.6E-08
1.8E-16
2.4E-15
Cs-137
1.3E-08
4.6E-09
6.2E-14
2.0E-13
Pb-210
6.9E-07
1.2E-06
1.5E-16
1.5E-15
Po-210
1.2E-06
3.3E-06
9.5E-19
3.0E-18
Ra-226
2.8E-07
3.5E-06
2.1E-13
6.5E-13
Ra-228
6.9E-07
1.6E-05
0
0
Th-229
6.1E-07
8.6E-05
2.9E-14
1.1E-13
Th-232
2.3E-07
1.1E-04
8.8E-18
5.9E-17
Pa-233
8.7E-10
3.9E-09
1.8E-14
6.7E-14
U-233
5.1E-08
3.5E-06
2.4E-17
1.1E-16
U-236
4.7E-08
8.7E-06
3.4E-18
3.2E-17
Np-237
1.1E-07
2.3E-05
1.4E-15
7.2E-15
Am-241
2.0E-07
4.2E-05
7.2E-16
5.5E-15
Pu-240
2.5E-07
5.0E-05
2.2E-18
2.9E-17
Pu-244
2.4E-07
4.7E-05
3.7E-14
1.2E-13
Cm-248
7.7E-07
1.5E-04
1.2E-18
2.0E-17
Cf-252
9.0E-08
2.0E-05
2.6E-18
3.0E-17
Notes
1. Values include effects of short-lived (half-life less than 25 days) daughters not
explicitly listed, assuming secular equilibrium at time of intake or exposure. A
list of short-lived daughters is given in Table L.3.
2. Data taken from [1].
132
3. Data taken from [2] assuming contamination to an infinite depth.
4. Data taken from [2].
References for Table D.5
[1]
ICRP (1996). Age dependent doses to members of the public from intake of
radionuclides: Part 5, Compilation of ingestion and inhalation dose
coefficients. ICRP Publication 72.Ann. ICRP 26 No. 1, Pergamon Press,
Oxford.
[2]
USEPA. Date Accessed: 16 September 2004. Online Database of Dose
Coefficients
from
Federal
Guidance
Report
No.
12.
http://www.ornl.gov/~wlj/fgr12tab.htm
[3]
UNSCEAR (2000).
Nations, New York.
D.2
Sources and Effects of Ionizing Radiation. United
ELEMENT-DEPENDENT DATA
TABLE D.6. EFFECTIVE DIFFUSION COEFFICIENTS (M2 Y-1)
Element
Co
Sr
I
Cs
Pb
Po
Ra
Th
Pa
U
Np
Pu
Am
Cm (7)
Cf (7)
Capsule
(1)
4E-2
4E-2
8E-2
8E-2
8E-2
8E-2
8E-2
1E-1
1E-1
1E-1
1E-1
1E-1
1E-1
1E-1
1E-1
Cement (3)
Undegraded Degraded
8E-5
8E-5
8E-5
8E-5
8E-5
8E-5
8E-5
8E-5
8E-5
8E-5
8E-5
8E-5
8E-5
8E-5
8E-5
4E-3
4E-3
4E-3
4E-3
4E-3
4E-3
4E-3
4E-3
4E-3
4E-3
4E-3
4E-3
4E-3
4E-3
4E-3
Fractured
System
(2)
9E-7 (4)
1E-5
3E-7
3E-5
1E-6 (6)
1E-6 (5)
1E-6
2E-7
1E-6
1E-6
1E-6
1E-6
1E-6
1E-6
1E-6
Flow Rate
Porous
System
(3)
2E-2
2E-2
2E-2
2E-2
2E-2
2E-2
2E-2
2E-2
2E-2
2E-2
2E-2
2E-2
2E-2
2E-2
2E-2
Notes
1. Data for free water diffusion from [1]. Used for capsule compartment since
capsule is assumed not to be backfilled (see Section 3.1.2).
133
2. Data from [2].
3. Data from [3].
4. Ni used as an analogue.
5. Pb used as an analogue.
6. Sn used as an analogue.
7. Am used as an analogue.
References for Table D.6
[1]
Little R H, van Blerk J, Walke R C and Bowden R A (2004). Generic PostClosure Safety Assessment and Derivation of Activity Limits for the Borehole
Disposal Concept. Quintessa Report QRS-1128A-6 v2.0, Quintessa Limited,
Henley-on-Thames, UK.
[2]
Andersson J (1999). SR 97 Data and Data Uncertainties.Compilation of Data
and Data Uncertainties for Radionuclide Transport Calculations.SKB
Technical Report TR-99-09, Swedish Nuclear Fuel and Waste Management
Company, Stockholm.
[3]
Savage, D. and M. J. Stenhouse (2002). SFR Vault Database. SKI Report R0253, Swedish Nuclear Power Inspectorate, Stockholm, Sweden.
134
TABLE D.7. SORPTION COEFFICIENTS (M3 KG-1)
Element
Capsule
(9)
Co
Sr
I
Cs
Pb
Po
Ra
Th
Pa
U
Np
Pu
Am
Cm
Cf
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
Cement
(6)
Undegraded Degraded
0.1
0.01
0.1
0.01
0.02
0.001
0.001
0.005
0.5
0.1
0.5
0.1
0.05
0.05
5
1
0.5
0.1
1/5 (5)
0.1/1 (5)
2/5 (5)
0.2/1 (5)
5
1
5
1
5
1
5 (1)
1 (1)
Saturated
Zone
(7)
0.1 (3)
0.005 (2)
0
0.05
0.1
0.1 (4)
0.5
1
1
1
1
1
5
5
5 (1)
Values for
Sandy Soil
(8)
0.06
0.013
0.001
0.27
0.27
0.15
0.49
3.0
0.54
0.033
0.0041
0.054
2.0
9.3 (10)
9.3 (1)
Notes
1. Cm used as an analogue.
2. In the absence of data in [4], it is assumed that Sr sorption values are an order of
magnitude lower than Cs values, consistent with the information given in [1] and
[3].
3. Pd used as an analogue.
4. Pb used as an analogue.
5. First value is for oxidising conditions, second is for reducing conditions.
6. Data from [2].
7. Data from [4] for sandstone with fresh type groundwater.
8. Data from [3].
135
9. Capsule is assumed not to be backfilled and so has no sorption properties (see
Section 3.1.2).
10. Data from [5].
References for Table D.7
[1]
Nagra (2002). Project Opalinus Clay: Models, Codes and Data for Safety
Assessment. Demonstration of Disposal Feasibility for Spent Fuel, Vitrified
High-level
Waste
and
Long-lived
Intermediate-level
Waste
(Entsorgungsnachweis). Nagra Technical Report 02-06.
[2]
Savage, D. and M. J. Stenhouse (2002). SFR Vault Database. SKI Report R0253, Swedish Nuclear Power Inspectorate, Stockholm, Sweden.
[3]
IAEA (2003). The Use of Safety Assessment in the Derivation of Activity
Limits for Disposal of Radioactive Waste to Near Surface Facilities. IAEATECDOC-1380, International Atomic Energy Agency, Vienna.
JNC (2000). H12: Project to Establish the Scientific and Technical Bais of HLW
Disposal in Japan. Supporting Report 3: Safety Assessment of the Geological
Disposal System. JNC Report JNC TN1410 2000-004, Japan Nuclear Cycle
Development Institute, Tokai, Japan.
[5]
IAEA (2010). Handbook of Parameter Values for the Prediction of
Radionuclide Transfer in Terrestrial and Freshwater Environments.
International Atomic Energy Agency Technical Report Series 472. Vienna,
Austria.
136
TABLE D.8. SOIL TO PLANT CONCENTRATION FACTORS (BQ KG-1 FRESH
WT/BQ KG-1 DRY SOIL) FOR CROPS
Element
Co
Sr
I
Cs
Pb
Po
Ra
Th
Pa
U
Np
Pu
Am
Cm (2)
Cf (2)
Root
Vegetables
(1)
3E-2
9E-2
1E-1
3E-2
1E-2
2E-4
4E-2
5E-4
4E-2
1E-3
1E-3
1E-3
1E-3
1E-3
1E-3
Green
Vegetables
(1)
3E-2
3E+0
1E-1
3E-2
1E-2
2E-4
4E-2
5E-4
4E-2
1E-3
1E-2
1E-4
1E-3
1E-3
1E-3
Notes
1. Data from [1].
2. Am used as an analogue.
References for Table D.8
IAEA (2003). The Use of Safety Assessment in the Derivation of Activity Limits
for Disposal of Radioactive Waste to Near Surface Facilities. IAEA-TECDOC1380, International Atomic Energy Agency, Vienna.
137
TABLE D.9: WEATHERING RATES (Y-1)
Element
Co
Sr
I
Cs
Pb
Po
Ra
Th
Pa
U
Np
Pu
Am
Cm
Cf (2)
Root
Vegetables
(1)
1.8E+1
1.8E+1
1.8E+1
1.8E+1
1.8E+1
1.8E+1
1.8E+1
1.8E+1
1.8E+1
1.8E+1
1.8E+1
1.8E+1
1.8E+1
1.8E+1
1.8E+1
Green
Vegetables
(1)
1.8E+1
1.8E+1
1.8E+1
1.8E+1
1.8E+1
1.8E+1
1.8E+1
1.8E+1
1.8E+1
1.8E+1
5.1E+1
5.1E+1
5.1E+1
1.8E+1
1.8E+1
Notes
1. Data taken from [1].
2. Cm used as an analogue.
References for Table D.9
[1]
Smith G M, Fearn H S, Smith K R, Davis J P and Klos R (1988). Assessment
of the radiological impact of disposal of solid radioactive waste at
Drigg.National Radiological Protection Board, NRPB-M148, Chilton, UK.
138
TABLE D.10: FRACTION OF ACTIVITY TRANSFERRED FROM EXTERNAL
TO INTERNAL PLANT SURFACES (-)
Element
Co
Sr
I
Cs
Pb
Po (2)
Ra
Th
Pa
U
Np
Pu
Am
Cm
Cf (3)
Root
Vegetables
(1)
1.7E-1
1.4E-1
7.4E-2
3.0E-1
2.2E-1
2.2E-1
9.9E-2
2.9E-1
2.9E-1
4.3E-2
2.9E-1
4.3E-2
2.9E-1
1.1E-1
1.1E-1
Green
Vegetables
(1)
1.8E-1
2.0E-1
6.1E-1
1.9E-1
2.2E-1
2.2E-1
1.8E-1
3.8E-2
4.5E-1
3.6E-1
4.5E-1
3.6E-1
2.8E-1
2.7E-1
2.7E-1
Notes
1. Data taken from for root vegetables and leafy vegetables [1].
2. Pb used as an analogue.
3. Cm used as an analogue.
References for Table D.10
[1]
Ashton J and Sumerling T J (1988). Biosphere Database for Assessments of
Radioactive Waste Disposals.UKDoE Report No. DoE/RW/88.083.
139
TABLE D.11. FOOD PREPARATION LOSSES (-)
Element
Co
Sr
Cs
Pb
Po
Ra
I
Th
Pa
U
Np
Pu
Am
Cm
Cf (3)
Root
Vegetables
(1)
0.0E+0
0.0E+0
0.0E+0
0.0E+0
0.0E+0
0.0E+0
0.0E+0
0.0E+0
0.0E+0
0.0E+0
0.0E+0
0.0E+0
0.0E+0
0.0E+0
0.0E+0
Green
Vegetables
(2)
9.0E-1
9.0E-1
9.0E-1
9.0E-1
9.0E-1
9.0E-1
9.0E-1
9.0E-1
9.0E-1
9.0E-1
9.0E-1
9.0E-1
9.0E-1
9.0E-1
9.0E-1
Notes
1. Data from [1].
2. Data from [2].
3. Cm used as an analogue.
References for Table D.11
[1]
Simmonds J R and Crick M J (1982). Transfer parameters for use in terrestrial
foodchain models.
National Radiological Protection Board, NRPB-M63,
Chilton, UK.
[2]
Smith G M, Fearn H S, Smith K R, Davis J P and Klos R (1988). Assessment
of the radiological impact of disposal of solid radioactive waste at
Drigg.National Radiological Protection Board, NRPB-M148, Chilton, UK.
140
TABLE D.12: TRANSFER COEFFICIENTS TO ANIMAL PRODUCE
Element
Co
Sr
I
Cs
Pb
Po
Ra
Th
Pa
U
Np
Pu
Am
Cm (2)
Cf (3)
Beef
(d kg-1 fresh
weight)
(1)
1.0E-2
8.0E-3
4.0E-2
5.0E-2
4.0E-4
5.0E-3
9.0E-4
2.7E-3
5.0E-5
3.0E-4
1.0E-3
1.0E-5
4.0E-5
9.8E-5
9.8E-5
Cow’s Milk
(d l-1)
(1)
(m3 kg-1 fresh
weight) (1)
3.0E-4
2.8E-3
1.0E-2
7.9E-3
3.0E-4
3.4E-4
1.3E-3
5.0E-6
5.0E-6
4.0E-4
5.0E-6
1.1E-6
1.5E-6
9.0E-6
9.0E-6
3E-1
6E-2
4E-2
2E+0
3E-1
5E-2
5E-2
1E-1
1E-2
1E-2
3E-2
3E-2
3E-2
3E-2
3E-2
Fish
Notes
1. Data from [1]. Data for fish only used for variant calculation that assumes
contaminated groundwater is used to supply a fish farm.
2. Data from [2].
3.
Cm used as an analogue.
References for Table D.12
[1]
IAEA (2003). The Use of Safety Assessment in the Derivation of Activity
Limits for Disposal of Radioactive Waste to Near Surface Facilities. IAEATECDOC-1380, International Atomic Energy Agency, Vienna.
[2]
Ashton J and Sumerling T J (1988). Biosphere Database for Assessments of
Radioactive Waste Disposals.UKDoE Report No. DoE/RW/88.083.
141
D.3
NEAR-FIELD ELEMENT-INDEPENDENT DATA
TABLE D.13. NEAR-FIELD LIQUID RELEASE AND FLOW DATA
Parameter
Hydraulic
conductivity
Total porosity
Grain density
Hydraulic gradient in
saturated zone (4)
Initial radius of
source
Corrosion/dissolution
rate of source (11)
Units
m y-1
kg m-3
-
Small and
Large
Capsule
1E+6 (6)
1.0E+0 (7)
1.0E+3 (8)
m
m y-1
Near-field
Cement
Undegraded
Degraded
3.2E-1 (1)
3.2E+2 (2)
1.0E-1 (3)
2.4E+3 (3)
0.05
2.5E-1 (3)
2.4E+3 (3)
5.5E-3 (small capsules) (9)
1.0E-2 (large capsule) (10)
1E-8
Notes
1. Based on data from [1] which gives range of 3.2E-3 to 3.2E-1 m y-1 for structural
cement.
2. Typical value for sand and gravel.
3. Data from [2].
4.
Data from [4] .
5. See Section 3.2.1.
6. Nominal value adopted to ensure flow in the near field is not limited by the
hydraulic conductivity of the capsule.
7. Capsule assumed to be void space.
8. Assumed to be the same as water.
9. Average radius of Co-60 sources to be disposed in small capsule (see Table 7).
10. Radius of Co-60 source to be disposed in the large capsule (see Table 7).
11. [3] gives a value of 1E-3 g m-2 d-1 (approx. 1E-7 m y-1) for glass at a temperature
of 60 °C and notes that the dissolution rate is about an order of magnitude lower
at 20 °C. It is conservatively assumed that the dissolution rate for ceramic will
be the same as for glass.
142
References for Table D.13
[1]
Nagra (1994). Report on Long-Term Safety of the L/ILW Repository at the
Wellenberg Site. Nagra Report NTB 94-06. Wettingen, Switzerland..
[2]
Allard, B., Höglund, L. O., and Skagius, K. (1991). Adsorption of
Radionuclides in Concrete, SKB Progress Report SKB/SFR 91-02, Swedish
Nuclear Fuel and Waste Management Company, Stockholm, Sweden.
[3]
JNC 2000. H12: Project to establish the scientific and technical basis for HLW
disposal in Japan. Second Progress Report on Research and Development for
the Geological Disposal of HLW in Japan. JNC Technical Report JNC
TN1410 2000-002.
[4]
Groundwater assessment report 1996, Water Research Institute, CSRI,
TABLE D.15. NEAR-FIELD TRANSPORT DATA
Compartment Type
Small Capsule (containing
source container)
1 (1)
Length in
Direction of
Flow
(m)
1.05E-2 (4)
Large Capsule (containing
source container)
1 (1)
2.40E-2 (4)
1.16E-2 (8)
3.25E-2
3.65E-1
Containment Barrier for
small capsule
Containment Barrier for
large capsule
Disposal Zone (horizontally
adjacent to capsule)
Disposal Zone (vertically
adjacent to capsule) (2)
Disturbed Zone (Backfill)
1 (1)
4.10E-2 (5)
6.83E-1 (8)
2.68E-2
2.15E+0
1 (1)
2.70E-2 (5)
3.49E-2 (8)
1.98E-2
1.10E-1
1 (1)
1.25E-2 (6)
5.32E+0 (8)
3.13E-2
1.67E+1
-
-
-
-
-
1 (1)
5.00E-2 (7)
9.88E+0 (8)
2.75E-1
3.10E+1
-
-
-
-
-
Disposal Zone (Plug) (3)
Number of
Compartments
Area
Perpendicular
to Flow
(m2)
8.32E-2 (8)
Diffusion
Length
(m)
(9)
2.58E-2
Area for
Diffusion
(m2)
(10)
2.61E-1
Notes
1. Flow is horizontal and so the flow path length through the disposals is equal
to the diameter of the borehole (0.26 m) (Section 3.1.2) and so one
compartment is sufficient to represent the flow path through each near-field
component.
143
2. The material within the disposal zone vertically adjacent to the capsule is not
explicitly modelled as it is assumed not to participate in the transport of
radionuclides.
3. Flow moves horizontally through the concrete backfill of the disturbed zone
between the casing and the borehole wall. The plug is not modelled since it is
assumed not to participate in the transport of radionuclides.
4. The outside radius of the small and large capsule is 10.5 mm and 24 mm,
respectively (Table 6). Flow moves horizontally through the capsules giving
a length in the direction of groundwater flow of 0.0105 m and 0.024 m for the
compartment used to represent the small and large capsule, respectively.
5. Flow moves horizontally through the containment barrier. The length of the
cement grout containment barrier in the direction of water flow is therefore
the same as its thickness, i.e. 41 mm for the containment barrier for the small
capsule and 27 mm for the containment barrier for the large capsule (Table 5).
6. Flow moves horizontally through the disposal zone. The length of the
disposal zone in the direction of water flow is therefore the same as the
distance between the disposal container and the borehole casing, i.e. 12.5 mm
(Section 3.1.2).
7. Flow moves horizontally through the backfill of the disturbed zone between
the casing and the borehole wall. The length of the disturbed zone in the
direction of water flow is therefore the same as the distance between the
borehole casing and the borehole wall, i.e. 50 mm (Section 3.1.2).
8. Flow moves horizontally through the borehole. Therefore, the area of each
compartment perpendicular to water flow is calculated using a formula based
on the depth of compartment multiplied by the diameter of compartment
9. Taken to be the distance between the mid points of the adjacent
compartments.
10. Equal to the circumference of the compartment multiplied by its depth.
144
TABLE D.16. TIMES FOR THE FAILURE OF THE PERFORMANCE OF THE
NEAR-FIELD COMPONENTS FOR THE DESIGN SCENARIO
Component
Failure Times (y, from time of disposal)
Aerobic Conditions
Anaerobic Conditions
Start of
Failure (3)
Totally
Failed (4)
Start of
Failure (3)
Totally
Failed (4)
Backfill Cement (1)
5.15E+02
5.64E+02
1.03E+03
1.13E+03
Stainless steel disposal container (2)
6.40E+02
6.40E+02
5.91E+03
5.91E+03
Containment Barrier (small capsule) (1)
8.49E+02
8.69E+02
6.32E+03
6.36E+03
Stainless steel capsule (small) (2)
9.55E+02
9.55E+02
8.75E+03
8.75E+03
Containment Barrier (large capsule) (1)
6.73E+02
6.76E+02
5.97E+03
5.98E+03
Stainless steel capsule (large) (2)
7.73E+02
7.73E+02
9.18E+03
9.18E+03
Notes
1. Data derived from Table H.1. (IAEA, 2008.)
2. Data taken from Table I.13. (IAEA, 2008.)
3. Represents start of degradation for cement grout (i.e. end of Stage 2/start of
Stage 3).
4. Represents end of degradation for cement grout (i.e. end of Stage 3/start of Stage
4).
145
TABLE D.17. TIMES FOR THE FAILURE OF THE PERFORMANCE OF THE
NEAR-FIELD COMPONENTS FOR THE DEFECT SCENARIO VARIANTS
Defect Scenario D1
Component
Failure Times (y, from time of disposal)
Aerobic Conditions
Start of Failure (8)
Anaerobic Co
Totally
Failed (9)
Start of
Failure (8)
Backfill Cement (1)
5.15E+02
5.64E+02
1.03E+03
Stainless steel disposal
container (2)
6.40E+02
6.40E+02
5.91E+03
Containment Barrier (small
capsule) (1)
8.49E+02
8.69E+02
6.32E+03
0
9.55E+02
0
6.73E+02
6.76E+02
5.97E+03
0
7.73E+02
0
Defective stainless steel
capsule (small) (5)
Containment Barrier (large
capsule) (1)
Defective stainless steel
capsule (large) (5)
146
F
Defect Scenario D2
Component
Failure Times (y, from time of disposal)
Aerobic Conditions
Anaerobic Conditions
Start of Failure
(8)
Totally
Failed (9)
Start of
Failure (8)
Totally
Failed (9)
Backfill Cement (1)
5.15E+02
5.64E+02
1.03E+03
1.13E+03
Failed stainless steel disposal
container (3)
0.00E+00
6.40E+02
0.00E+00
5.91E+03
Containment Barrier (small
capsule) (1)
5.15E+02
5.64E+02
4.19E+03
4.59E+03
Stainless steel capsule (small)
in failed disposal container (4)
6.30E+02
6.30E+02
6.89E+03
6.89E+03
Containment Barrier (large
capsule) (1)
5.15E+02
5.64E+02
1.03E+03
1.13E+03
Stainless steel capsule (large)
in failed disposal container (4)
6.30E+02
6.30E+02
4.31E+03
4.31E+03
147
Defective Scenario D3
Component
Failure Times (y, from time of disposal)
Aerobic Conditions
Anaerobic Conditions
Start of
Failure (8)
Totally
Failed (9)
Start of
Failure (8)
Totally
Failed (9)
Backfill Cement (6)
2.58E+01
4.94E+01
5.15E+01
9.95E+01
Stainless steel disposal container (2)
1.38E+02
1.38E+02
4.90E+03
4.90E+03
Containment Barrier (small capsule) (6)
1.48E+02
1.58E+02
4.92E+03
4.94E+03
Stainless steel capsule (small) (2)
2.51E+02
2.51E+02
7.34E+03
7.34E+03
Containment Barrier (large capsule) (6)
1.39E+02
1.41E+02
4.90E+03
4.90E+03
Stainless steel capsule (large) (2)
2.40E+02
2.40E+02
8.10E+03
8.10E+03
148
Defective Scenario D4
Component
Failure Times (y, from time of disposal)
Aerobic Conditions
Anaerobic Conditions
Start of
Failure (8)
Totally
Failed (9)
Start of
Failure (8)
Totally
Failed (9)
5.15E+02
5.64E+02
1.03E+03
1.13E+03
Failed stainless steel disposal container
(7)
0
6.40E+02
0
5.91E+03
Containment Barrier (small capsule) in
failed disposal container (1)
5.15E+02
5.64E+02
4.19E+03
4.59E+03
Failed stainless steel capsule (small) in
failed disposal container (7)
0
9.55E+02
0
8.75E+03
Containment Barrier (large capsule) in
failed disposal container (1)
5.15E+02
5.64E+02
1.03E+03
1.13E+03
Failed stainless steel capsule (large) in
failed disposal container (2)
0
7.73E+02
0
9.18E+03
Backfill Cement (1)
149
Notes
1. Data derived from Table H.1. of IAEA, 2008.
2. Data taken from Table I.13. of IAEA, 2008.
3. Assumes that disposal container has a defective weld.
4. Early failure of disposal container only affects failure time of capsule in that
container.
5. Assumes that one capsule has a defective weld. It is assumed that the defective
weld allows access to 10% of the waste once the disposal container has failed. It
is assumed that, once the defective capsule fails, there is a ramp up to 100% of
the waste being available. Note that although the defective capsule is assumed to
have failed at t=0, no releases of radionuclides occur until the waste container is
breached.
6. Data derived from Table H.2 of (IAEA, 2008).
7. Assumes that one capsule has a defective weld and this is contained in a disposal
container that also has a defective weld. It is assumed that the defective weld in
the capsule allows access to 10% of the waste. It is assumed that, once the
defective capsule fails, there is a ramp up to 100% of the waste being available.
8. Represents start of degradation for cement grout (i.e. end of Stage 2/start of
Stage 3).
9. Represents end of degradation for cement grout (i.e. end of Stage 3/start of Stage
4).
150
D.4
GEOSPHERE ELEMENT-INDEPENDENT DATA
TABLE D.18. GEOSPHERE FLOW DATA
Parameter
Saturated Geosphere
Units
Porous System
m y-1
7.3E+1
Hydraulic gradient (1)
-
5E-2
Total porosity (1)
-
1.5E-1
Degree of saturation (3)
-
1E+0
kg m-3
2.65E+3
Hydraulic conductivity (1)
Grain density (4)
Fractured System
Fracture
7.3E+1
Matrix
Fracture
Matrix
Fracture
Matrix
Fracture
Matrix
Fracture
Matrix
5E-2
1.5E-1
5E-3 (2)
1E+0
1E+0
2.65E+3
2.65E+3
151
Notes
1. Data is taken from Section 3.2.3. It is assumed in the fractured system that the matrix does not contribute to flow and so hydraulic
conductivity and hydraulic gradient values do not need to be specified.
2. For fractured system assume matrix porosity to be 5E-3 (consistent with [1]).
3. By definition, the degree of saturation in the saturated zone is unity.
4. Grain density of quartz assumed.
References for Table D.18
[1]
Andersson J (1999). SR 97 Data and Data Uncertainties.Compilation of Data and Data Uncertainties for Radionuclide Transport
Calculations.SKB Technical Report TR-99-09, Swedish Nuclear Fuel and Waste Management Company, Stockholm.
TABLE D.20. GEOSPHERE TRANSPORT DATA FOR DISPOSAL IN THE SATURATED ZONE
System
Parameter
Number of compartments
between the disposal borehole
and the abstraction borehole
Length of each compartment in
direction of water flow (3)
Area of each compartment
perpendicular to water flow (4)
Diffusion length between
adjacent compartments (6)
Area over which diffusion
occurs (8)
Units
Porous System
-
5 (1)
m
20
m2
38
m
20
m2
38
Fractured System
Fracture
5 (1)
Matrix
5 (2)
Fracture
Matrix
Fracture
Matrix
Fracture
Matrix
Fracture
Matrix (m2 m-3)
20
20 (2)
38
- (5)
20
0.02 (7)
38
1 (9)
Notes
1. For an advection dominated system, the number of compartments should equal the Peclet number divided by two (see [1]). Peclet number is
equal to the distance from the disposal borehole to the abstraction borehole (100 m ) divide by the longitudinal dispersion length (assumed to
be 10% (see [2]) of the distance from the disposal borehole to the abstraction borehole).
2. Assumes that each fracture compartment has an associated matrix compartment between which there is a diffusive flux.
153
3. The distance to the abstraction borehole from the disposal borehole (100 m) divided by the number of compartments in the saturated zone
(5).
4. It is assumed that the radionuclides enter the saturated zone over the entire length of the disposal zone considered in the AMBER model (38
m) and then are dispersed transverse to the direction of flow. The degree of transverse dispersion is assumed to be 1% of the distance to the
abstraction borehole, consistent with [3], resulting in a plume cross-sectional area of 50 m2..
5. It is assumed that there is no flow water in the matrix.
6. Assumed to be equal to the length of each compartment.
7. Value taken from [4] for rock matrix depth.
8. Assumed to equal to the area of each compartment perpendicular to water flow.
9. Assumes that transverse diffusion occurs from/to fracture into/from rock matrix (see Equation 19 and Equation 20). Values represent the
flow wetted surface area per unit volume of rock and are based on data given in [4] taking into the assumed degree of saturation (see Table
D.18).
References for Table D.20
[1]
Penfold, J.S.S., R.H. Little, P.C. Robinson, and D. Savage. 2002. Improved Safety Assessment Modelling of Immobilised LLW Packages
for Disposal. Ontario Power Generation Technical Report 05386-REP-03469.3-10002-R00. Toronto, Ontario.
[2]
IAEA (2004). Improvement of Safety Assessment Methodologies for Near Surface Disposal Facilities. Volume I: Review and
Enhancement of Safety Assessment Approaches and Tools. IAEA-ISAM-1, International Atomic Energy Agency, Vienna.
[3]
Neuman, S.P. (1990). Universal Scaling of Hydraulic Conductivities and Dispersivities in Geological Media, Water Resources Research,
Vol. 26, No. 8, pp. 1749-1758.
154
[4]
Andersson J (1999). SR 97 Data and Data Uncertainties.Compilation of Data and Data Uncertainties for Radionuclide Transport
Calculations.SKB Technical Report TR-99-09, Swedish Nuclear Fuel and Waste Management Company, Stockholm.
TABLE D.21. FRACTION OF WATER DEMAND SUPPLIED BY CONTAMINATED WATER FOR DIFFERENT GEOSPHERES AND
CALCULATION CASES
Drinking Water Only Calculation Case
All Other Cases
Location of Disposal Zone
Saturated Zone (1)
Location of Disposal Zone
Saturated Zone (2)
Porous system
1.00E+0
5.22E-1
Fractured
system
1.00E+0
5.22E-1
Geosphere
Note
1. The minimum value of unity and the result of dividing the water flux in which the contaminated plume is mixed (139 m3 y-1– derived by
multiplying the cross-sectional area of the plume (Table D.20) by the hydraulic gradient and hydraulic conductivity (Table D.19)), by the
assumed drinking water abstraction rate (2.92 m3 y-1) (2 l d-1 per person).
2. The minimum value of unity and the result of dividing the water flux in which the contaminated plume is mixed (139 m3 y-1), by the assumed
water abstraction rate (266 m3 y-1) (Table D.22).
155
D.5
BIOSPHERE ELEMENT-INDEPENDENT DATA
TABLE D.22.BIOSPHERE COMPARTMENT PARAMETERS AND PROCESSES
Parameter
Depth
Length
Width
Total porosity
Degree of saturation
Grain density
Percolation rate
Inhalable dust concentration
Erosion rate
Volume of irrigation water that
reaches the soil
Volume of non-irrigation water
plus irrigation water intercepted
by crops
Site Dweller
(9)
Units
Farmer
(8)
m
m
m
kg m-3
m y-1
kg m-3
m y-1
Surface Soil
2.5E-1 (1)
3.51E+1 (2)
1E+1 (2)
3E-1 (1)
3.3E-1 (3)
2.65E+3 (4)
5E-2 (3)
2E-8 (1)
1E-3 (5)
Surface Soil
2.5E-1 (1)
3.51E+1 (2)
1E+1 (2)
3E-1(1)
3.3E-1 (3)
2.65E+3 (4)
5E-2 (3)
2E-8 (1)
1E-3 (5)
m3 y-1
71 (6)
- (10)
m3 y-1
195 (7)
- (10)
156
Notes
1. Data taken from [1].
2. An area of 351 m2 is required to grow root and green vegetables to meet the
assumed demand of an exposure group of four people (Table D.23), assuming the
yields given in Table D.24. Assuming a nominal width of 10 m, the length is
therefore 35.1 m.
3. Consistent with [2].
4. Grain density of quartz.
5. See Section 3.3.1.
6. Value derived by multiplying the depth of irrigation water applied to root and
green vegetables (Table D.24), the area of root and green vegetables required to
meet the assumed demand of an exposure group of four people (Table D.23)
(assuming the yields given in Table D.24), and unity minus the interception
fraction for irrigation water (Table D.24).
7. Value derived by summing the volume of water intercepted by crops and the
volume of water required by cows and humans. The volume of water intercepted
by crops is calculated by multiplying the depth of irrigation water applied to root
and green vegetables (Table D.24), the area of root and green vegetables required
to meet the assumed demand of an exposure group of four people (Table D.23)
(assuming the yields given in Table D.24), and the interception fraction for
irrigation water (Table D.24). The volume of water required by cows is
calculated by multiplying the number of cows (i.e. four) required to meet the
assumed meet and milk demands of an exposure group of four people (Table
D.23) by the annual water consumption rate of cows (derived from the daily rate
given in Table D.24). The volume of water required by humans is calculated by
multiplying the number of humans in the exposure group (i.e. four) by the annual
water consumption rate of humans (Table D.23).
8. Exposed via the liquid release resulting from corrosion of the disposal container
and waste capsule.
9.
Exposed via the solid release resulting from erosion of the cover zone.
10. Does not use contaminated water.
References for Table D.22
[1]
IAEA (2003). The Use of Safety Assessment in the Derivation of Activity
Limits for Disposal of Radioactive Waste to Near Surface Facilities. IAEATECDOC-1380, International Atomic Energy Agency, Vienna.
157
[2]
IAEA (2008). Generic Post-closure Safety Assessment for Borehole Disposal
of Disused Sealed Sources. IAEA Draft Safety Report (Draft 0.8), International
Atomic Energy Agency, Vienna.
[3]
Little R H, van Blerk J, Walke R C and Bowden R A (2004). Generic PostClosure Safety Assessment and Derivation of Activity Limits for the Borehole
Disposal Concept. Quintessa Report QRS-1128A-6 v2.0, Quintessa Limited,
Henley-on-Thames, UK.
[4]
BNFL (2002). Drigg Post-Closure Safety Case. British Nuclear Fuels plc,
Sellafield.
TABLE D.23. HUMAN BEHAVIOUR PARAMETERS
Exposure Mechanism
Contaminated drinking water
Contaminated root
vegetables
Contaminated green
vegetables
Contaminated beef
Ingestion
Contaminated cow’s milk
Contaminated soil
Contaminated fish
Contaminated outdoor air
Contaminated indoor air
Time spent on contaminated
soil
Time spent in contaminated
building
Time spent in contaminated
water
Inhalation
Occupancy
Units
m3 y-1
kg fw
y-1
kg fw
y-1
kg fw
y-1
kg fw
y-1
kg fw
h-1
kg fw
y-1
m3 h-1
m3 h-1
Farmer
0.73 (1)
Exposure Group
House
Site Dweller
Dweller
(6)
-
235 (2)
-
235 (2)
62 (2)
-
62 (2)
95 (2)
-
300 (2)
-
-
1.5E-5 (3)
-
-
6.9 (4)
-
-
1 (2)
-
0.75 (2)
1 (2)
-
h y-1
2192 (2)
-
2192 (2)
h y-1
-
6575 (2)
-
h y-1
365 (5)
-
-
Notes
1.
Assumes water consumption rate of 2 l/d.
2.
Data taken from [1].
3.
Data taken from [2] assuming that the annual value quoted in [2] results from an
exposure to contaminated soil of 8 hours per day.
158
4.
Data taken from [3]. Ingestion of fish only considered for the variant
calculation that assumes contaminated groundwater is used to supply a fish
farm.
5.
Assumes 1 h d-1. Only considered for the variant calculation that assumes
contaminated groundwater is used for bathing.
References for Table D.23
[1]
IAEA (2003). Only exposed through growing crops on contaminated
soil due to erosion of closure zone.
The Use of Safety Assessment in the Derivation of Activity Limits for Disposal of
Radioactive Waste to Near Surface Facilities. IAEA-TECDOC-1380,
International Atomic Energy Agency, Vienna.
[2]
Yu C, Loureiro C, Cheng J-J, Jones L G, Wang Y Y, Chia Y P and Faillace E
(1993). Data Collection Handbook to Support Modelling the Impacts of
Radioactive Material in Soil.Argonne National Laboratory, Report
ANL/EA15-8.
[3]
IAEA (2003). “Reference Biospheres” for Solid Radioactive Waste Disposal:
Report of BIOMASS Theme 1 of the BIOsphere Modelling and ASSessment
(BIOMASS Programme). IAEA-BIOMASS-6, International Atomic Energy
Agency, Vienna.
TABLE D.24. NON-ELEMENT DEPENDENT PLANT PARAMETERS
Parameter
Soil contamination of crop
Yield of crop
Depth of irrigation water
applied to crop
Interception fraction for
irrigation water
Time interval between
irrigation and harvesting
kg dw soil/kg fw crop (1)
kg fw m-2 y-1 (2)
Root
Vegetables
1.5E-4
3.5E+0
Green
Vegetables
1.0E-4
3.0E+0
m y-1 (2) (3)
3.0E-1
3.0E-1
- (2) (3)
3.3E-1
3.3E-1
y (3) (4)
4.0E-2
2.E-2
Units
Notes
1. Data taken from [1].
2. Data taken from [2].
159
3. Irrigation of crop with contaminated water only considered for the liquid release
calculation cases.
4. Data taken from [3].
References for Table D.24
[1]
Little R H, van Blerk J, Walke R C and Bowden R A (2004). Generic PostClosure Safety Assessment and Derivation of Activity Limits for the Borehole
Disposal Concept. Quintessa Report QRS-1128A-6 v2.0, Quintessa Limited,
Henley-on-Thames, UK.
[2]
IAEA (2003). The Use of Safety Assessment in the Derivation of Activity
Limits for Disposal of Radioactive Waste to Near Surface Facilities. IAEATECDOC-1380, International Atomic Energy Agency, Vienna.
[3]
IAEA (2003). “Reference Biospheres” for Solid Radioactive Waste Disposal:
Report of BIOMASS Theme 1 of the BIOsphere Modelling and ASSessment
(BIOMASS Programme). IAEA-BIOMASS-6, International Atomic Energy
Agency, Vienna.
TABLE D.25. NON-ELEMENT DEPENDENT ANIMAL PARAMETERS
Parameter
Units
Cows
Consumption of water
m3 d-1
6E-2 (1)
Notes
1. Data taken from [1].
References for Table D.25
[1]
IAEA (2003). The Use of Safety Assessment in the Derivation of Activity
Limits for Disposal of Radioactive Waste to Near Surface Facilities. IAEATECDOC-1380, International Atomic Energy Agency, Vienna.
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