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Session 3c – ILW IAEA-CN-242
International Conference on the Safety of Radioactive waste Management
SESSION 3c
Disposal of
Intermediate Level Waste
Session 3c – ILW IAEA-CN-242
2
ORAL PRESENTATIONS
No. ID Presenter Title of Paper Page
03c – 01 64 R. Nakabayashi
Japan
Development of Methodology for Probabilistic
Safety Assessment of Long Term Radioactive
Waste Disposal
4
03c – 02 91 S. Konopaskova
Czech Republic
Waste Acceptance Criteria Development for
Different Low and Intermediate Level Waste
(LILW) Disposal Systems
8
03c – 03 92 E. Andersson
Sweden
Assessment of Human Intrusion and Future
Human Actions – Example from the Swedish
Low and Intermediate Level Waste Repository
SFR
13
03c – 04 97 A. Carter
United Kingdom
Data Management to Support a Post-Closure
Safety Case for Higher Activity Wastes
17
03c – 05 193 H. Arlt
United States of
America
Greater-Than-Class C Low Level Radioactive
Waste Characteristics and Disposal Aspects
21
03c – 06 135 J.-M. Hoorelbeke
France
Implementation of a Graded Approach in
Radioactive Waste Management in France
26
Session 3c – ILW IAEA-CN-242
3
POSTER PRESENTATIONS
No. ID Presenter Title of Paper Page
03c – 07 80 K. Källström
Sweden
Methodology and Results for the Safety
Assessment for Low and Intermediate level
Waste Repository (SFR) in Sweden
30
03c – 08 95 A. Glindkamp
Germany
Implementation of Requirements on the
Chemical Toxicity of Nuclear Waste at a
Repository
35
03c – 09 102 M. Nepeypivo
(A. Talitskaya)
Russian Federation
Safety Assessment as an Instrument for Waste
Acceptance Criteria Derivation
38
03c – 10 128 B. Samwer
Germany
Konrad Repository – Evaluation on the Safety
Requirements according to the State of the Art
of Science and Technology
43
Session 3c – ILW IAEA-CN-242
4
03c – 01 / ID 64. Disposal of Intermediate Level Waste
DEVELOPMENT OF METHODOLOGY FOR PROBABILISTIC SAFETY
ASSESSMENT OF LONG-TERM RADIOACTIVE WASTE DISPOSAL
R. Nakabayashi, D. Sugiyama
Central Research Institute of Electric Power Industry (CRIEPI), Tokyo, Japan
E-mail contact of main author: [email protected]
Abstract. This paper discusses a methodology for probabilistic safety assessment for long-term radioactive
waste disposal considering both the epistemic and aleatory uncertainties included in the safety assessment. This
methodology can be used to demonstrate compliance with dose criteria and be helpful for the optimization of
radiation protection in a waste disposal programme. In addition, the applicability of the probabilistic approach is
demonstrated by illustrating a safety assessment for a radioactive waste disposal facility in Japan.
Key Words: Probabilistic safety assessment; uncertainty; dose criteria; optimization
1. Introduction
For the protection of people after the closure of a disposal facility, the disposal facility has to
be designed so as not to exceed the dose constraint that is used as a dose criterion, and
radiation protection is required to be optimized [1]. In disposal of long-lived radioactive
waste, safety assessment must take into consideration not only aleatory uncertainties but also
epistemic uncertainties. An aleatory uncertainty originates from the inherent heterogeneity or
diversity of data (e.g., the fracture permeability of host rock), and an epistemic uncertainty is
due to lack of knowledge (e.g., the degradation time of an engineered barrier). In this paper,
we briefly review a methodology for probabilistic safety assessment considering both
epistemic and aleatory uncertainties [2]. This method was developed to determine compliance
with the dose criterion of 0.3 mSv/year and to provide useful material for the optimization of
radiation protection. In addition, the applicability of the probabilistic safety assessment is
demonstrated by illustrating a safety assessment for a radioactive waste disposal facility in
Japan.
2. Framework of the methodology of probabilistic safety assessment
2.1.Probabilistic dose assessment
The procedure of the probabilistic dose assessment [2] is briefly described as follows.
(1) Aleatory and epistemic uncertainties are quantified as a probability distribution by
applying a statistical process to measured data and eliciting expert judgments,
respectively.
(2) Probabilistic dose assessment is carried out in consideration of both the epistemic and
aleatory uncertainties by using the radionuclide migration program with a Monte Carlo
simulation.
(3) The cumulative distribution function (CDF) of the maximum annual dose of a certain
radionuclide is calculated in the dose assessment. The probability density function (PDF)
is also calculated by kernel density estimation.
Session 3c – ILW IAEA-CN-242
5
2.2.Demonstration of compliance with a dose criterion
It is possible to demonstrate the compliance with a dose criterion by comparing the 95th
percentile of the CDF with 0.3 mSv/year and analogically adopting the concept of the
‘representative person’ [3]. The ICRP recommends that the representative person should be
defined such that the probability is less than about 5% that a person drawn at random from
the population will receive a greater dose in a probabilistic safety assessment. This indicates
that the vast majority of the population is protected from the radiation when the 95th
percentile of the dose distribution incorporating the uncertainties involved is less than the
dose criterion. By demonstrating that the 95th
percentile of the dose distribution, obtained by
probabilistic dose assessment in consideration of the uncertainties associated with long-term
radioactive waste disposal, is less than 0.3 mSv/year, the aim of protection of the public is
achieved. In Figs. 1(a) and 1(b), the maximum annual dose C is adopted as the assessment
result for comparison with the dose criterion of 0.3 mSv/year.
2.3.Optimization of radiation protection
In the optimization of radiation protection through compliance with the dose criterion of 0.3
mSv/year, not only the 95th
percentile of the CDF but also the mode of the PDF should be
reduced to as low as reasonably achievable while taking economic and social factors into
account. The mode of the PDF is the most likely dose that the public will be exposed to,
which is derived from the most likely behavior of the disposal system. Efforts to reduce the
most likely dose lead to the increased safety of a waste disposal facility. If more than one
option is capable of providing the required level of safety (i.e., the 95th
percentile of the CDF
is less than 0.3 mSv/year), then other factors, which are economic and social, also have to be
considered [1]. We propose that the mode of the PDF is one of the most important factors in
addition to the 95th
percentile of the CDF for the optimization of radiation protection.
If a regulatory body sets out a dose criterion for the most likely behavior of a disposal system,
the determination of the compliance with the dose criterion should performed conservatively.
In this case, the larger of the modal value of the PDF and the 50th
percentile of the CDF can
be employed to meet the dose criterion as discussed in a previous paper [2]. In Fig. 1(a),
where dose B (mode of PDF) is greater than dose A (50th
percentile of CDF), dose B is
adopted as the assessment result for comparison with the dose criterion. In Fig. 1(b), where
dose B (mode of PDF) is less than dose A (50th
percentile of CDF), dose A is adopted as the
assessment result for comparison with the dose criterion.
FIG. 1. Concept of the approach to determine the compliance with dose criteria [2].
Session 3c – ILW IAEA-CN-242
6
3. Application of probabilistic safety assessment
This section outlines how we carry out a safety assessment for radioactive waste disposal in
Japan using the probabilistic approach. Note that this is only an example used to discuss the
applicability of the probabilistic approach.
The regulatory body in Japan requires applicants to demonstrate that their dose assessment
results are less than the dose criteria assigned for each scenario in consideration of the
uncertainties. The dose criteria of likely and less-likely scenarios, which are classified in
terms of their likelihood of occurrence based on a disaggregated dose/probability approach,
are 0.01 and 0.3 mSv/year, respectively [4]. The purpose of conducting the safety assessment
of the likely scenario is to evaluate whether the basic design of the disposal system has been
considered to minimize the effects of radiation on the public (i.e., less than 0.01 mSv/year)
under normally expected scenarios. The purpose of the less-likely scenario is to check
whether the doses based on the scenario are below the dose criterion of 0.3 mSv/year, even
when taking into account uncertainties that are less likely but may have a significant effect in
the safety assessment. This example presents the assessment of compliance with the likely
and less-likely scenarios in consideration of the epistemic and aleatory uncertainties
associated with a sub-surface disposal system.
We consider the dose assessment model for exposure pathways in groundwater migration in a
sub-surface disposal system and deal with the epistemic uncertainty concerning the
degradation times of the engineered barriers and the aleatory uncertainty concerning the
permeability coefficient of the host rock. The engineered barriers are composed of
cementitious or bentonite materials. 14
C (4.38×1015
Bq) is instantaneously released from
radioactive waste in this model.
3.1. Quantification of epistemic uncertainty and aleatory uncertainty
The epistemic uncertainty concerning the degradation time of each barrier was expressed as a
subjective probability distributions on the basis of expert judgment (Fig. 2). The aleatory
uncertainty concerning the permeability coefficient of the host rock was expressed as a log-
normal distribution by applying a statistical process to measured values (Fig. 2). For details
of the quantification, refer to Nakabayashi and Sugiyama (2016) [2].
FIG. 2. Probability distributions for the epistemic uncertainty concerning the degradation time of a
cementitious barrier (a) and bentonite barrier (b), and aleatory uncertainty concerning the
permeability coefficient of the host rock (c) used in the safety assessment [2].
Session 3c – ILW IAEA-CN-242
7
FIG. 3. PDF and CDF indicating the maximum annual dose of 14
C [2].
3.2. Determination of compliance with the likely and less-likely scenarios
The PDF and CDF of the maximum annual dose of 14
C were obtained from the probabilistic
safety assessment (Fig. 3). In this section, we illustrate how to demonstrate the assessment
results in compliance with the stepwise dose criteria (0.01 and 0.3 mSv/year) of likely and
less-likely scenarios. The mode of the PDF was 0.0025 mSv/year, whereas the 50th
percentile
of the CDF was 0.0029 mSv/year, i.e., the 50th
percentile was larger than the mode. In this
case, the 50th
percentile is adopted as the assessment result for comparison with the dose
criteria in the likely scenario. The 95th
percentile of the CDF is adopted as the assessment
result for comparison with the dose criteria of 0.3 mSv/year in the less-likely scenario.
4. Conclusion
A probabilistic safety assessment considering epistemic and aleatory uncertainties has been
proposed to determine the compliance with a dose constraint of 0.3 mSv/year. This
methodology can estimate the mode of the PDF, which is the most likely dose that the public
will be exposed to. For the optimization of radiation protection, it is important to strive to
reduce the 95th
percentile of the CDF and the mode of the PDF to as low as reasonably
achievable while taking economic and social factors into account.
REFERENCES
[1] International Atomic Energy Agency, Disposal of Radioactive Waste, IAEA Safety
Standards Series No. SSR-5, IAEA, Vienna (2011).
[2] Nakabayashi, R., Sugiyama, D., Development of Methodology of Probabilistic Safety
Assessment for Radioactive Waste Disposal in Consideration of Epistemic Uncertainty
and Aleatory Uncertainty, Journal of Nuclear Science and Technology, Taylor & Francis
(2016).
[3] International Commission on Radiological Protection, Assessing Dose of the
Representative Person for the Purpose of Radiation Protection of the Public and the
Optimisation of Radiological Protection, Publication 101, Pergamon Press, Oxford and
New York (2006).
[4] Nuclear Safety Commission of Japan, Basic Policy for Safety Regulations Concerning
Land Disposal of Low-Level Waste (Interim Report), NSC, Tokyo (2007).
Session 3c – ILW IAEA-CN-242
8
03c – 02 / ID 91. Disposal of Intermediate Level Waste
WASTE ACCEPTANCE CRITERIA DEVELOPMENT FOR DIFFERENT LILW
DISPOSAL SYSTEMS
S. Konopaskova, D. Lukin, I. Zadakova
Radioactive waste repository authority (SURAO), Praha, Czech Republic
E-mail contact of main author: [email protected]
Abstract. In the Czech Republic, there are operated two types of repositories: near surface for disposal of low
level waste from NPPs, and repositories for institutional waste, specified as low and intermediate level waste;
these are located underground, in former mines of different types. New legislation after 1997 and optimized
conditions for final waste form characterization lead to improvement of WAC derivation methods by the means
of safety assessment and supported their variety.
Key Words: waste acceptance criteria, subsurface repository, repository for low and
intermediate level waste, safety assessment
1. Introduction
This paper describes the procedure of waste acceptance criteria (WAC) development, applied
for various types of operated radioactive waste repositories in the Czech Republic. Safety
related criteria are derived using the results of safety assessment, considering waste streams,
barriers system, and position of the repository in the host structure. Special considerations are
included evaluating hydrogeological conditions of the host structure and accessible
biosphere. Differences of repositories lead to differences in WAC, as it is presented below.
2. WAC for disposal systems in the Czech Republic
In the Czech Republic, there are operated two types of radioactive waste repositories:
Subsurface disposal of waste from nuclear power plants
Disposal of low and intermediate level institutional waste in mine cavities, some
tenths of meters below surface
WAC defined for individual repositories differ in extent, qualitative expression and
quantitative parameters thanks to specific approach to their derivation, considering different
project, operational and environmental conditions of the repositories.
2.1. Operated repositories and their types
The overview of repositories is specified in Table 1.
Session 3c – ILW IAEA-CN-242
9
TABLE I: MARGINS FOR YOUR MANUSCRIPT.
Site Type Volume Waste streams Matrix Host rock
Dukovany,
1993 -
subsurface 55 000 m3
Waste from NPPs Bitumen,
geopolymer
Crystalline
Richard,
1964 -
LILW 17 000 m3
Institutional waste,
artificial radionuclides
Cement Limestone
mine
Bratrství,
1973 -
LILW 1 200 m3 Institutional waste,
natural radionuclides
Cement Uranium mine
FIG. 1. Dukovany repository – a subsurface vault system
FIG. 2. Richard and Bratrství repositoris – underground disposal systems
2.2.WAC structure
WAC are structured according to safety requirements, technical restrictions and
administrative requirements defined by law.
Safety related criteria guarantee the compliance with qualitative and quantitative objectives
of nuclear safety and radiation protection. These criteria are derived from the results of safety
assessment, namely:
Total activity of radionuclides in the repository
Volume activity of radionuclides in different waste forms
Session 3c – ILW IAEA-CN-242
10
Leachability of the final waste form
Activity of radionuclides in non-solidified waste
Stress resistance of final waste form solidified by cement and/or geopolymer
Dose rate on the surface of waste package
Technical restrictions are done by repository construction and include:
Water presence in drainage system
Weight of waste package
Integrity and structural stability of the waste package
Administrative and formal restrictions guarantee the compliance with nuclear and
environmental legislation:
Presence of free liquids, pyrophoric and toxic substances, complexing and
microbiological agents
Waste tracking system and passport
2.3. Derivation of safety related WAC
In the procedure of safety related WAC, there are more aspects that can lead to differences in
WAC specifications, done by:
Composition of the radionuclide vector
Final waste form and packages properties
Repository construction
Depth of the repository below surface
Hydrogeological conditions of the host structure, and
Probable use of the land in communication points
Generally, there is defined a set of scenarios supporting safety case, i. e. operational scenarios
and long term scenarios that should help to evaluate probable radiation effects during
operations and after repository closure.
Normal evolution scenario is used to define the capacity of the site - volume of the waste as
well as its total activity. Scenario components are site specific. For subsurface system, direct
infiltration of rainwater and advective flow through disposal units are considered
immediately after institutional control period. For underground system, the infiltration is
controlled by diffusion and by inflow to fractures in near field and advective flow starts much
later thanks to final waste form and filling stability. Safety function of host structure is
strongly affected by hydrogeology system as a part of transport pathway in all types of
repositories.
Alternative scenarios are used to evaluate disposal system performance by deviations from
projected performance. For near surface repository, bathtubbing is considered; in mine
systems, possible outflow of contaminated mine water is taken into account.
Intrusion scenarios are the base for limiting volume activities in the final waste form. In
subsurface system, evaluation of on site residence and working activities on the site after
institutional control are considered. For underground repositories, there is evaluated a contact
with waste as a consequence of drilling activities.
Limits of dose rates in the controlled area and on the waste packages are derived by means of
radiation protection. In addition, radon intake has to be considered in underground
repositories.
Session 3c – ILW IAEA-CN-242
11
Potential emergency situations could lead to non acceptable doses of workers, but the
evaluation of accidents is not considered in formulating criteria for radioactive content of the
waste.
2.4. Quantitative comparison of safety related criteria
The limits of total and volume activities have been compared for various disposal systems, as
it is indicated in FIG. 1. In spite of the fact that the composition of radioactive waste and its
activities are not identical in waste from NPPs and in institutional waste, there are some
radionuclides present in both types of waste. The results of safety cases lead to lower
permitted activities and activity concentration in the vault system and the capacity.
Underground repositories provide higher capacity for activity of short lived radionuclides as
well as for long lived radionuclides, thanks to sophisticated stabilization system, better
hydrogeological conditions, as well as to lower potential for inadvertent intrusion.
FIG. 3 .Limits of activity in vault and underground systems
3. Conclusions
Normal evolution scenario is limiting for both total and volume activities in mine
repositories, Higher capacity of mine system is shown thanks to better engineered barriers
performance and/or lower probability of intrusion.
EBS system is more efficient in mine systems – barrier can assure diffusion driven transport
for longer periods of time even in the case that the number of fractures in near field is
relatively high.
In subsurface system, increasing of life time of barriers leads to higher doses in intrusion and
on site scenarios as a consequence of negligible decrease of activity of long lived
radionuclides during institutional control period. Volume activities in subsurface system have
to be strongly limited, to the values of very low level waste.
REFERENCES
[1] INTERNATIONAL ATOMIC ENERGY AGENCY, Derivation of activity limits for
the disposal of radioactive waste in near surface disposal facilities, IAEA-TECDOC-
1380, Vienna (2003).
[2] KONOPASKOVA, S., et al., “Safety report of Dukovany repository”, SÚRAO, Praha
(2012).
Session 3c – ILW IAEA-CN-242
12
[3] KONOPASKOVA, S., et al., “Safety report of Richard repository”, SÚRAO, Praha
(2014).
[4] MILICKY, M. et al., Hydrogeological model of the Dukovany site, ProGeo 2012
[5] MILICKY, M. et al., Hydrogeological model of the Richard site, ProGeo 2013
[6] Regulation SONS No. 307/2007 Coll. on radiation protection
Session 3c – ILW IAEA-CN-242
13
03c – 03 / ID 92. Disposal of Intermediate Level Waste
ASSESSMENT OF HUMAN INTRUSION AND FUTURE HUMAN ACTIONS -
EXAMPLE FROM THE SWEDISH LOW AND INTERMEDIATE LEVEL WASTE
REPOSITORY SFR
E. Andersson1, T. Hjerpe
2, G. Smith
3, K. Källström
1, L. Morén
1, K. Skagius
1
1Swedish Nuclear Fuel and Waste Management Co., Sweden
2Facilia AB, Sweden,
3GMS Abingdon Ltd, UK
E-mail contact of main author: [email protected]
Abstract. The strategy commonly adopted in the disposal of solid radioactive waste is to contain the waste so
that it is kept away from the accessible biosphere by means of underground disposal. The intention is to isolate
the waste from man and the biosphere for a sufficiently long time to allow radioactive decay to significantly
reduce the radiation hazard. However, the potential exposure to radioactive material following intrusion is an
inescapable consequence of the deposition of the radioactive waste in a repository. There is an international
consensus that future human actions (FHA) resulting in disruption of the disposal facility must be considered in
the safety assessment as part of the safety case for a radioactive waste repository. However, although there are
some general recommendations concerning assessment of radioactive waste disposal, there is no over-arching
international methodological guide on how to perform FHA assessments. There is an ongoing project at IAEA
on handling inadvertent human intrusion (HIDRA). The Swedish Nuclear Fuel and Waste Management
Company (SKB) is taking part in the HIDRA project for human intrusion but also analyse the broader concept
FHA.
SKB has performed several analyses of FHA (including human intrusion by drilling) for both the existing
repository for low- and intermediate level waste (SFR) situated at 60-120 m depth and for the planned repository
for spent nuclear fuel to be situated at approximately 500 m depth. The SKB methodology to assess FHA
includes FEP-analysis, identification of stylised scenarios and qualitative and quantitative evaluation of the
stylised scenarios.
In December 2014, SKB submitted an application to the Swedish Radiation Protection Authority (SSM) to
extend the existing repository for low- and intermediate level waste (SFR). The planned extension includes 6
additional rock caverns to be placed at 120 m depth. The safety case for the application included an assessment
of FHA for both the existing part of the repository and for the planned extension. The methodology used and
major results of the FHA analysis are presented. In addition, examples are given of adjustments to general
recommendations that were needed to address FHA issues relevant to the assessment needs for this specific
assessment.
Key Words: Human intrusion, future human actions, waste disposal, safety assessment
1. Introduction
There is a long-standing international consensus that future human actions (FHA) and human
intrusion (HI) resulting in some disruption to the repository must be considered in safety
assessments as part of a safety case for a radioactive waste repository [1, 2, 3]. However,
there is no over-arching international guide on how to incorporate FHA in assessments. IAEA
has an ongoing project, ‘Human Intrusion in the context of Disposal of RAdioactive waste’
(HIDRA) to develop and test a methodology [4]. There are also other international projects
where experiences of handling HI in different countries have been shared e.g. [5] and further
commentary provided [6]. Depending on site specific and repository specific conditions as
well as regulatory and local stakeholder considerations, different aspects of FHA may need to
Session 3c – ILW IAEA-CN-242
14
be considered. SKB has addressed FHA in safety assessment since the late 90’s. In this paper,
the methodology [7] used in the assessment of FHA for the low and intermediate level waste
(L/ILW) repository SFR is described.
2. General recommendations by international projects
Although an international FHA methodology is not available, there are useful
recommendations in documents like those mentioned above. Below is a list of some typical
examples and a note when they have not been followed in the SFR assessment [7].
Select a site away from natural resources in order to minimize likelihood for intrusion.
Only consider inadvertent intrusion, i.e. actions carried out when the location of the
repository is unknown, its purpose forgotten or the consequences of the actions are
unknown. Current society cannot be required to protect future societies from their
own intentional and planned activities if they are aware of the consequences.
A common approach to societal conditions is to use current conditions, both regarding
human behavior and technological development. Sites my change due to e.g. climate
change, then current data from sites with similar conditions may be used in the
assessment. In the area where SFR is situated, land uplift leads to areas currently
covered by sea to be situated below dry land. This has been addressed in the FHA
analysis.
Avoid quantitative use of probabilities because it is difficult to justify assigning a
number to the probability of specific FHA. Nevertheless, some quantitative
consideration of probabilities of such events is considered in the FHA assessment for
SFR.
Instead of trying to identify every possible feature, event and process (FEP) and
analyze all possible FHA, it is recommended to use a few stylized scenarios to
illustrate the range of consequences if they were to occur. However, a FEP-list is a
good tool to identify a consolidated set of relevant scenarios and this approach has
been used in the SFR assessment.
3. Relevant features of the repository SFR in Sweden
SFR is an existing repository for L/ILW situated below the sea floor in the Baltic Sea. The
sea is currently a barrier for HI but due to the ongoing post-glacial land uplift SFR will be
situated below land in the future and then HI will be possible. The assessment needed to
consider these altered future conditions at the site even though the geosphere remains an
effective barrier. SFR consist of 4 rock vaults and one silo situated between 60-120 m depth
in granitoid rock. In 2014, SKB applied to extend the repository with 6 rock vaults and filed a
safety assessment including assessment of FHA [7].
4. Methodology with examples from the SFR assessment
In the assessment of FHA for SFR, a step-wise methodology was used (Fig. 1).
5. Analysis of FEPs
A FEP-list was produced by first identifying safety relevant factors and then identifying
actions (FEPs) related to FHA that could negatively affect these safety factors. The audited
FEP-list proved to be a good tool for generating stylized scenarios and in communication
Session 3c – ILW IAEA-CN-242
15
with the public. Our experience suggests that treating future human actions that the public are
concerned could pose a hazard to future generations in similar manner to the FEPs in the
main risk assessment can help to build confidence in the safety case.
FIG. 1. Overview of the stepwise methodology used for handling FHA at SKB.
6. Scenarios and calculation cases
Based on the FEP-list, a consolidated set of stylized scenarios was identified, taking into
account stakeholder interests. In this consolidation all FEPs were covered by at least one
scenario unless there were effective and documented arguments that the FEP would not affect
the robustness of the safety case. The scenarios evaluated were:
Drilling scenario including four separate calculation cases
o Exposure due to utilizing the drilling hole as a well
o Exposure to on-site crew during the drilling
o Exposure during construction on drilling detritus landfill
o Exposure due to cultivation on drilling detritus landfill
Underground construction scenario
Scenario with mine in the vicinity of the repository scenario
7. Evaluation of results and use of probabilities
In Sweden, drinking wells commonly reach a depth of 60 m and so exposure due to utilizing
an intrusion well cannot be ruled out. The drilling scenario calculation with exposure due to
utilizing water from an intrusion well was included in the main risk assessment for which
compliance with the regulatory risk criterion of 10-6
/y (nominally comparable to 14 µSv/y)
needed to be assessed. The doses for the well scenario were relatively high, up to 4.5 mSv.
However, the footprint area of the repository is small and the likelihood of a well in this area
Session 3c – ILW IAEA-CN-242
16
is very low. Thus, it was deemed appropriate to assess intrusion wells as a less probable
scenario and assign probabilities. For the majority of the analyzed period, the main scenario
made up the largest risk but around 3000 AD, drilling directly into one of the rock vaults
accounted for the highest risk. The total risk summed over all scenarios was below the risk
criterion of 10-6
/y for the entire assessment period of 100 000 years.
For other drilling scenarios (drilling personnel, construction worker and farmer), the doses
was always low, at most 0.25 mSv. This is well below the ICRP ranges of reference levels
indicative of system robustness (ICRP, 2013). Use of these reference levels is another way of
addressing the generally low likelihood of intrusion without explicit consideration of the
probability. The FHA scenarios, mining in the area and water management work, were
evaluated qualitatively. FHA were considered already in siting and these scenarios were
determined to be unlikely and to have little effect on the repository.
8. Conclusion
An international consensus on how to assess FHA would be very welcome and useful. It is
hoped that SKB work, shared though mechanisms such as HIDRA is a useful contribution to
development of suitable guidance. However, there will always be relevant site, waste type
and repository design factors to take into account when conducting specific safety
assessments, alongside local stakeholder interests and national regulatory requirements.
REFERENCES
[1] NEA, Risks Associated with Human Intrusion at Radioactive Waste Disposal Sites.
Proceedings of an NEA Workshop. Nuclear Energy Agency, Paris (1989).
[2] INTERNATIONAL ATOMIC ENERGY AGENCY, Disposal of Radioactive Waste,
Specific Safety Requirements, IAEA Safety Standards Series No. SSR-5, IAEA, Vienna
(2011).
[3] INTERNATIONAL COMMISSION ON RADIOLOGICAL PROTECTION,
Radiological Protection in Geological Disposal of Long-lived Solid Radioactive Waste,
ICRP Publication 122 (2013).
[4] SEITZ, R., et al., "Role of Human Intrusion in Decision-Making for Radioactive Waste
Disposal - Results of the IAEA HIDRA Project - 16287," Proceedings from the WM2016
Conference, March 6 - 10, 2016, Phoenix, AZ, 201 (2016).
[5] BAILEY L., et al., PAMINA Performance assessment methodologies in application to
guide development of the safety case. European Handbook of state-of-the-art of the safety
assessments of geological repositories – Part 1. Deliverable 1.1.3 (Ch. 9), European
Commission, (2011).
[6] SMITH G.M., et al., Human Intruder Dose Assessment for Deep Geological Disposal.
Report prepared under the BIOPROTA international programme. Available at
www.bioprota.org (2012).
[7] SKB, 2014, Safety analysis for SFR Long term safety. Main report for the safety
assessment SR-PSU. Technical report TR-14-01, Swedish Nuclear Fuel and Waste
Management Co, Stockholm.
Session 3c – ILW IAEA-CN-242
17
03c – 04 / ID 97. Disposal of Intermediate Level Waste
DATA MANAGEMENT TO SUPPORT A POST-CLOSURE SAFETY CASE FOR
HIGHER ACTIVITY WASTES
A.J. Carter, L.E.F. Bailey
Radioactive Waste Management Ltd., Building 587, Curie Avenue, Harwell Campus, Didcot,
Oxfordshire, OX11 0RH, UK
E-mail contact of main author: [email protected]
Abstract. In this paper we describe how RWM has developed its approach to data and model management,
starting from a set of formal ‘aims and principles’; and promoted a culture in which these can operate
effectively. In the course of this development, a number of innovative systems and tools have also been
produced which facilitate the storage and use of data. These will be described, as will lessons learned during the
roll out to date.
Key Words: Data management, Model management, Safety case production.
1. Introduction
The United Kingdom (UK) is committed to the safe management and disposal of higher
activity radioactive waste. This will be carried out through the interim storage of radioactive
waste packages prior to their final disposal in a deep geological disposal facility [1].
Radioactive Waste Management Ltd (RWM) is responsible for the delivery of such a facility,
and maintains a generic Environmental Safety Case (ESC) [2] for UK wastes while a site is
identified through a siting process in partnership with local communities and government.
Following the production of the 2013 UK Radioactive Waste Inventory, the generic ESC is
being updated to take into account changes to waste inventory and packaging, and to reflect
developments in scientific understanding which has resulted from new research since 2010,
when the previous generic ESC was published. The generic ESC is supported by a generic
post-closure safety assessment [3] which presents illustrative calculations to support RWM’s
confidence that a safety case, consistent with the regulatory risk guidance level and other
stakeholder expectations, could be produced in UK-relevant geologies. The probabilistic
computer models underlying these calculations are referred to as total system models (TSMs)
as they contain high-level representations of the total system, that is the wasteform, container,
engineered barrier system, geosphere and biosphere. Significant quantities of input data are
required for these models covering multiple disciplines across RWM’s programme.
In the period since 2010, RWM has undertaken a formal project to review and update its
procedures relating to data and model management. The project has received strong support
from RWM’s Executive team and has resulted in significant improvements in the way data
and models are documented, managed and used across the company. These have now been
applied in the production of the TSMs introduced above and have thereby helped to ensure
the quality and traceability of calculations which are used to support the safety case.
In this paper we describe how RWM has developed its approach to data and model
management, starting from a set of formal ‘aims and principles’; and promoted a culture in
which these can operate effectively. In the course of this development, a number of
innovative systems and tools have also been produced which facilitate the storage and use of
data. These will be described, as will lessons learned during the roll out to date.
Session 3c – ILW IAEA-CN-242
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2. Data Management
An initial step in the data management project involved writing a high-level Policy and
Principles document which recognises the company’s reliance on data and sets out a vision
for the management of its data. This is followed by a set of five principles:
Data and information management is part of everyone’s role;
Data are an asset;
Data and information quality will be assured at source and maintained;
Data and information will be accessible; and
Data and information integrity and security will be assured.
Each principle is used to formally derive a set of implications, and these in turn are used to
inform the underlying data management procedure and define the requirements of its
supporting systems and tools. As an example, the principle that ‘data and information will be
accessible’ implies that each dataset should be made available quickly after its acquisition,
that a means should be provided for staff to discover the dataset, and that a mechanism
should be provided for staff to obtain or access the dataset. The procedure addresses these by
providing flow charts to describe how to register a new dataset, record the use of a dataset
and retire a dataset. The principle also implies a technical requirement to store data using a
sensible file format and attach appropriate metadata to facilitate search. Similarly the
principle that ‘data and information integrity and security will be assured’ implies
requirements for access control, backup, business continuity/recovery arrangements, the
creation of audit trails as datasets are periodically maintained or updated, and the use of
storage locations which minimise the potential for decay or corruption of data.
The accountability for each dataset lies with a senior member of staff from an appropriate
department, known as the data owner. Data owners are accountable to the organisation for the
security, integrity, quality and availability of their data, including making adequate provision
for its long-term care and ensuring it is managed in line with the data procedure. Ownership
of data at a senior level with the organisation helps to reinforce the importance of data
management while ensuring appropriate oversight of data, and its use in business decision
making, at a strategic level. Data owners may delegate their day to day responsibility for a
dataset (for example fielding queries from users) to a data steward, typically a member of
staff within their department, but still retain overall accountability for the dataset.
To support the proper characterisation of a dataset, a number of attributes have been defined:
Characterisation of uncertainty: the degree to which uncertainty, variability and
precision in the data are understood and represented in the dataset;
Provenance: the presence of information within the dataset to describe its source,
underlying assumptions and methods of production; and
Limitations of applicability: the degree to which the dataset addresses the entire scope
of the domain, for example spatial or temporal, together with its internal consistency
(for example in values, terminology, production methodology or in the adoption of
standards).
Each attribute must be populated by a suitably qualified and experienced person who
understands the dataset – ideally at the time of the creation of the dataset – and should be
stored with the dataset. An additional attribute has also been defined to support the proper use
of a dataset:
Relevance: the degree to which the dataset meets the needs of a particular use.
Again this attribute must be populated by suitably qualified and experienced people, typically
recording the agreement between a data owner or steward, who understands each dataset
Session 3c – ILW IAEA-CN-242
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(together with any other datasets which may be relevant), and a data user, who understands
the use the data will be put to (for example a model technical owner, introduced in Section 3,
who understands the underlying conceptual model). In practice agreement from multiple
owners or stewards is likely to be required for a model which covers multiple disciplines.
To achieve its mission to build and operate a geological disposal facility, RWM must
demonstrate that it has the required nuclear safety and environmental competencies for each
stage of process. To support this, a competence management system has been introduced in
order to define and assess whether its staff are suitably qualified and experienced in each
area. The availability of staff with appropriate competencies will also be reviewed as the
siting process progresses and is also used to inform recruitment needs. A competence panel,
chaired by the chief scientific advisor, is used to determine whether a member of staff is
currently regarded as competent in a given area, using an agreed list of requirements for that
area. Evidence of competency, including a list of formal qualifications, skills and experience
are used to make this judgement and are captured via the completion of a competence
assessment form. Any training and development needs which arise from the competence
panel are captured on the development plan for the member of staff, and are reviewed as part
of the performance management process at six monthly intervals. Introduction of a
competence management system has also helped RWM management to identify business
risks, for example skill shortages or key skills which reside with a single member of staff. At
least one of the data owner or data steward must be regarded as competent in the appropriate
technical area for a dataset and this is confirmed as part of the approval process.
Two electronic forms have been developed to support the new data management process,
named a data definition form (DDF) and a data use form (DUF). The DDF is used to define a
dataset, and to record the data quality attributes relating to uncertainty, provenance and
applicability described above. A DUF is used to identify a data need (for example
radionuclide half-lives for the total system model) together with the dataset which will be
used. The decision on the dataset to use takes place through agreement between a data owner
or steward and the data user, as noted above, and is recorded via completion of the relevance
attribute on the DUF, where any caveats or risks on the data of this data for this purpose are
also documented. The DUF for a model which covers multiple disciplines is likely to
reference several DDFs, each potentially with a different data owner or steward, and the
relevance attribute would need to be populated for each DDF used.
The forms are created and edited using a custom .NET based application which saves the
DDF or DUF in an XML compliant format. The application includes the ability to refer to
reference documents stored within the company knowledge base and allows numerical data to
be either directly entered into the form or referenced to an external file (with a secure
checksum used to ensure integrity). Deterministic values and probability distribution
functions (PDFs) are supported and the application is dimensionally aware, so that suppliers
of data may enter each physical quantity in the system of units most appropriate to their field.
An application programming interface (API) has been created so that each DDF (or DUF), or
table of data within a DDF, or individual value may be accessed from code, with filters
available to export the data to GoldSim (used for the total system model) and Microsoft
Excel. This helps to remove transcription errors, and the code is also able to convert each
item of data to a specified system of units thereby removing unit conversion errors. An
additional export filter has recently been produced which is able to inject data into a template
document using the Office Open XML format (for example a .docx file used by Microsoft
Word), thereby allowing automated population of a data report. While the use of code to
populate models and reports carries an initial overhead, subsequent updates are made
considerably easily.
Session 3c – ILW IAEA-CN-242
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3. Model Management
A key component of the updated approach to model management requires the production of a
model register together with a model risk assessment and quality plan (MRAQP) for each
model. The model register, which is easily accessible over the web from the intranet home
page, identifies each model which is used by the company, together with a model senior
responsible owner [4] (MSRO) and model technical owner (MTO), analogous to the data
owner and data steward introduced above. The register also provides hyperlinks to the
MRAQP’s and storage locations for the version-controlled models themselves. ‘Write’ access
to each folder is restricted to the MSRO and MTO to prevent model users changing master
copies, and users are required to consult the register each time they make use of a model to
ensure that they are aware of any updates.
The MRAQP asks the MSRO to describe what the model does, provide a commentary on the
key uncertainties in model results and formally identify the uses of the model (for example by
listing the products results from the model feed into). On the basis of the key uncertainties
and model uses, an overall risk assessment is produced for the model, including an
identification of whether the model is ‘business critical’, and this is used to inform the overall
quality plan for the model. The overall quality plan identifies the level of model verification,
validation and benchmarking required, together with the arrangements for model planning,
version control, design, build and sign-off. It should also give guidance on the extent to
which uncertainties and caveats need to highlighted when presenting model outputs, together
with any other risk mitigations which have been identified. The MRAQP is intended to be a
live document which evolves with the model. It is formally approved by the MSRO and then
made readily available to users of the model or its results.
4. Conclusions
RWM is currently updating its generic ESC to take into account changes to waste inventory
and packaging, and to reflect developments in scientific understanding which have resulted
from new research since 2010, when the previous generic ESC was published. Since this time
the company has reviewed and significantly improved its procedures relating to data and
model management, as well as develop systems and tools to help support this process. Within
this paper, a discussion has been provided to explain how these improvements were
developed, together with a description of the key elements within the system. The updated
total system model which provides illustrative calculations for the new ESC is fully
compliant with the requirements of this system. In future RWM intends to investigate the use
of electronic signatures and electronic workflow to further improve the system.
REFERENCES
[1] DEPARTMENT OF ENERGY & CLIMATE CHANGE, Implementing Geological
Disposal, URN 14D/235, July 2014.
[2] RADIOACTIVE WASTE MANAGEMENT., Generic Environmental Safety Case
Main Report, DSSC/203/01, In publication.
[3] RADIOACTIVE WASTE MANAGEMENT., Generic Post-closure Safety Assessment,
DSSC/321/01, In publication.
[4] HM TREASURY, Review of Quality Assurance of Government Analytical Models:
Final Report, Nick Macpherson, March 2013.
Session 3c – ILW IAEA-CN-242
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03c – 05 / ID 193. Disposal of Intermediate Low Level Waste
GREATER-THAN-CLASS C LOW-LEVEL RADIOACTIVE WASTE
CHARACTERISTICS AND DISPOSAL ASPECTS
H. Arlt, T. Brimfield, C. Grossman
United States Nuclear Regulatory Commission (NRC), Washington, D.C., United States
E-mail contact of main author: [email protected]
Abstract. United States regulations (Part 61.55 of Title 10 of the Code of Federal Regulations, Waste
Classification) divides Low-level Radioactive Waste (LLRW) into three classes based on the concentration
levels of certain long-lived and short-lived radionuclides. The three waste classes are Class A, B, and C with
Class C having the higher concentration and/or more long-lived radionuclides than the other two classes.
Greater-Than-Class C (GTCC) waste is LLRW that exceeds the Class C concentration limits and is generally
not acceptable for near-surface disposal. GTCC LLRW corresponds to the low- and intermediate level waste
classes identified in the International Atomic Energy Agency’s Classification of Radioactive Waste General
Safety Guide No. 1. The disposal of GTCC LLRW is associated with greater challenges than other classes of
LLRW due to various waste streams having higher specific activities and higher concentrations of long-lived
radioactivity. The U.S. Department of Energy is responsible for the disposal for GTCC LLRW. The paper
contains insights from a qualitative examination of individual GTCC LLRW streams, disposal methods,
disposal environments, exposure scenarios including by means of inadvertent intrusion and groundwater
transport, and the significant interrelationships between these disposal aspects.
Key Words: Greater-Than-Class C, Low-level Radioactive Waste, waste types, disposal
methods
1. Introduction and Background
United States (U.S.) regulations (Part 61.55 of Title 10 of the Code of Federal Regulations,
Waste Classification, or 10 CFR 61.55) were promulgated to ensure the safe land disposal of
low-level radioactive waste (LLRW). The 10 CFR 61.2 definition of LLRW is based on the
exclusion of other waste streams, i.e., LLRW is defined as “radioactive waste not classified
as high-level radioactive waste, transuranic waste, spent nuclear fuel, or byproduct material
as defined in paragraphs (2), (3), and (4) of the definition of Byproduct material” set forth in
10 CFR 20.1003. The regulations divide LLRW into Class A, B, and C where Class A is the
least radiologically hazardous of the three classes and Class C has the higher concentration
levels of certain long-lived and short-lived radionuclides. LLRW that exceeds the Class C
limit, referred to as Greater-Than-Class C (GTCC) waste, is identified as generally not
acceptable for near-surface disposal although U.S. regulation at 10 CFR 61.55(a)(2)(iv)
allows for disposal in a near-surface facility if approved by the U.S. Nuclear Regulatory
Commission (NRC). The U.S. Department of Energy (DOE) is the responsible U.S. federal
agency for disposing of GTCC LLRW. At this time, there is no disposal capability for GTCC
LLRW; however, the DOE has published their final environmental impact statement [1]
which is an important step in the process towards obtaining GTCC LLWR disposal
capability.
A qualitative examination has provided a more comprehensive understanding of the risks
associated with site characteristics and disposal methods when considering GTCC LLRW
disposal [2]. This paper presents a summary of that examination including aspects that need
to be considered for disposal and also discusses disposal challenges under different
environmental settings and exposure scenarios. Specifically, insights were gained by
Session 3c – ILW IAEA-CN-242
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examining individual GTCC LLRW streams, disposal methods, disposal environments,
exposure scenarios, and the interrelationships between these disposal aspects. Performance
assessments of potential disposal sites containing GTCC LLRW would need to examine these
aspects of disposal. The majority of the information and the data on inventory in this paper
was obtained from Ref. [1] and [3].
2. Intermediate-Level Waste and GTCC LLWR
Intermediate-level waste (ILW) is defined by the International Atomic Energy Agency
(IAEA) [4] as waste that contains long-lived radionuclides in quantities that need a greater
degree of containment and isolation from the biosphere than is provided by near-surface
disposal. ILW contains waste with activity levels above clearance levels as described in Ref.
[5]. ILW may contain alpha-emitting radionuclides that will not decay to a level of activity
concentration acceptable for near-surface disposal during institutional controls. In addition,
ILW does not contain levels of activity concentration high enough to generate significant
quantities of heat by the radioactive decay process and has thermal output that is less than
2 kW·m–3
[6]. ILW is generally recommended for disposal at a depth of between a few tens
to a few hundreds of meters.
The radionuclides, activity concentrations, physical and chemical properties and other
characteristics of GTCC LLRW vary considerably and will influence the appropriate
regulatory approach to the disposal of GTCC LLRW including the depth at which it will be
disposed and a disposal site’s dependence on engineered barriers. However, the majority of
GTCC LLRW is more clearly aligned with IAEA’s definition of ILW than it is with the other
IAEA waste classes due to the properties of long-lived radionuclides in the GTCC LLRW
and that GTCC LLRW generally produces less than 2 kW·m–3
thermal output.
3. Characteristics of GTCC LLRW and Wasteforms
DOE has categorized three GTCC LLRW types: activated metals, sealed sources, and GTCC
Other Waste [1]. GTCC LLRW consisting of activated metals can include irradiated metal
components from reactors such as core shrouds, support plates, and core barrels, as well as
filters and resins from reactor operations and decommissioning [7]. Sealed sources are the
second type of GTCC LLRW and are used at hospitals, medical schools, research facilities,
industries, and universities. A third waste type that is not an activated metal or a sealed
source is referred to as GTCC “Other Waste” based on its differing radionuclides and
concentration levels and can consist of contaminated equipment, rubble, scrap metal, filters,
soil, and solidified sludges [7]. The total stored and projected volume of GTCC LLRW in the
U.S. will be approximately 8,800 m3 (311,000 ft
3) and the projected activity of that waste by
2083 will be 5.92 x 106 TBq (160 MCi) [1].
Activated metals are activated by neutron exposure and have a higher activity level than the
other GTCC LLRW types. Activated metals waste can be subdivided into two categories:
routinely generated activated metal and decommissioning activated metals [7]. The neutron
activation products expected to be most dominant in activated metals at the time of disposal
are C-14, Mn-54, Fe-55, Ni-59, Co-60, Ni-63, Mo-93, and Nb-94. Lower concentrations of
some fission products such as Sr-90, Tc-99, I-129, and Cs-137 and various isotopes of
plutonium are also expected to be present on these materials as surface contamination [8].
The projected total volume of activated metal waste is 2,000 m3 (71,000 ft
3) with 5.9 x 10
6
TBq (160 MCi) of activity, although most of the commercial reactors are not scheduled to
undergo decommissioning for several decades [1].
Session 3c – ILW IAEA-CN-242
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Sealed sources are generally small and the radionuclides are generally enclosed in capsules
made, with very few exceptions, of stainless steel, titanium, platinum or other inert metals
and encompassing several physical forms, including ceramic oxides, salts, or metals. Sealed
sources include Cs-137 irradiators that, unlike the smaller sealed sources, are larger than the
standard 208 liter (55 gallon) drum and would be disposed of individually [1]. GTCC sealed
sources may contain one of many radionuclides including Cs-137, Pu-238, Pu-239, Am-241,
and Cu-244, and their activities can range from 4.07 x 10-4
TBq (0.011Ci) to 1.5 x 105 TBq
(4.1 MCi). The projected volume of GTCC commercial sealed sources is 2,900 m3
(102,000 ft3) [1]. In many cases, the volume includes the device as well as the source since it
may be expeditious to dispose of the device and source as a unit.
The total stored and projected GTCC Other Waste activity is 1.98 x 104 TBq (0.53 MCi) and
relatively small compared to activated metals and sealed sources, although GTCC Other
Waste has a large variety of radionuclides, includes some very long-lived actinide isotopes,
and comprises the largest volume of the three waste types with 3,900 m3 (138,000 ft
3) [1]. A
wide spectrum of radionuclides can be present in this waste type with the isotopes of various
actinides (e.g., uranium, neptunium, plutonium, americium, and curium) being of higher
concern with regard to long-term waste management [8]. GTCC Other Waste generated from
routine operations includes contaminated clothing, floor sweepings, paper and plastic while
decommissioning waste can include building, piping, hardware, and equipment debris.
4. Disposal Aspects
Currently, disposal of LLRW that is not GTCC occurs near the surface with favorable
topographic and geological characteristics and/or with engineered barriers and other features
that impede or limit the eventual release of radionuclides from those facilities. The goal of
disposal is to isolate or limit the release of radioactive waste to the environment for hundreds
to thousands of years. For disposal sites with favorable geological and climatic
characteristics, natural barriers will reduce the number of engineered barriers needed to slow
contaminant release into the groundwater and atmosphere. However, modern disposal
practices can include multi-barrier systems that employ both natural and man-made
engineered barriers. The disposal methods chosen for GTCC LLRW disposal will be critical
to ensuring long-term safety.
In this paper, the lower boundary of near-surface disposal sites is considered to lie 30 m
(circa 100 ft) below the local topographic low point since 30 m is considered the maximum
depth of excavation for the foundations of tall buildings [5]. No generally agreed upon value
to define intermediate depth exists. However, most of the literature uses depths that start at
the near-surface lower boundary and include depths as deep as 100 - 150 m (300 to 500 ft)
under the surface [6] [9]. The disposal methods discussed [1] include disposal in concrete
structures or in trenches near the surface, disposal in borehole and shafts at intermediate
depths, and disposal in a deep geologic repository.
For the deep geologic repository disposal method, disposal sites could be located in semiarid
and arid environments as well as humid environments. However, sites with humid
environments would need to be designed for favorable saturated disposal conditions or be
located in hydrogeological settings that allow relatively dry conditions below water table
elevations (e.g., salt deposits, very compact clay layers, dry bedrock). Intrusion could only
occur if a borehole was drilled very deep. Assuming an inadvertent intruder-driller exposure
scenario was plausible, technical bases and assumptions concerning the degradation of the
stabilizing agent (e.g., grout) and the corrosion rate of metals would be important. For
exposure scenarios involving groundwater transport offsite, performance assessment results
Session 3c – ILW IAEA-CN-242
24
[10] indicate that activated metals would contribute more than the other GTCC LLRW types
during any activity concentration release.
For a borehole disposal method at an intermediate depth, disposal sites with humid
environments would be suitable if waste is designed to be disposed in a saturated
environment or if hydrogeological setting allowed waste to be placed in deeper, yet dry,
disposal sites. If an inadvertent intruder-driller exposure scenario is considered plausible,
depth and size of the waste package (e.g., sealed sources) would need to be factored into the
plausibility. Assuming such an exposure scenario is plausible, technical bases and
assumptions concerning the degradation of the stabilizing agent, concrete, and outer canister
of sealed sources would be significant as would the corrosion of activated metals. For GTCC
Other Waste, area concentration limits during disposal may be possible. For the groundwater
transport exposure scenario, the borehole disposal showed the lowest peak dose in
comparison with the trench and vault disposal methods [1].
For a concrete structural containment disposal method and the trench disposal method at the
near surface, disposal sites with humid environments could be suitable; however, additional
engineering controls would be required to address the higher precipitation and potential
erosion rates at these sites. Ref. [1] trench design includes a 5 m (16 ft) minimum cover that
would be deeper than most building construction sites to limit the potential for inadvertent
intruder exposure scenarios. Increased infiltration rates (relative to an arid site) would make
grout, concrete and metal degradation rates especially important. If the GTCC Other Waste
from the West Valley Site in the State of New York was included in the calculations for
groundwater transport at humid sites, GTTC Other Waste would be the main contributor to
peak dose due to the readily soluble nature of this waste type in comparison to activated
metals and seal sources. The radionuclides contributing to peak dose include C-14, I-129,
uranium, and transuranic radionuclides including isotopes of plutonium and americium. For
waste disposal in above-ground concrete structures, the vulnerability to erosional processes
increases potentially allowing more infiltration to occur as the overlying material becomes
less thick and root zones move closer to the barriers. For the semiarid to arid sites, peak doses
were lower [1]. For intruder-driller exposure scenarios, degradation assumptions are again
significant since a driller most likely would not drill through a large intact metallic
wasteform, but would be more likely to drill through a degraded wasteform or degraded
concrete barrier.
5. Summary
A qualitative examination of the challenges associated with GTCC LLRW disposal have
shown how the interrelationships between different disposal site characteristics and the
diverse GTCC LLRW types would make it difficult to regulate the disposal of such waste
within a prescriptive, generic framework: guidelines that may apply to one GTCC LLRW
type may not apply to the other; a disposal method that allows adequate performance in one
environmental setting performs poorly in another. Currently, NRC staff, as the U.S. regulator
for GTCC LLRW, is carrying out a quantitative examination of the different GTCC LLRW
types in context with the many disposal aspects. If NRC staff concludes, as a result of its
analysis, that that some or all GTCC LLRW is potentially suitable for near-surface disposal
with or without special processing, design, or site suitability conditions, NRC staff would
proceed with the development of a proposed rule to include disposal criteria for licensing the
disposal of such waste.
Session 3c – ILW IAEA-CN-242
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REFERENCES
[1] UNITED STATES DEPARTMENT OF ENERGY, Final Environmental Impact
Statement for the Disposal of Greater-Than-Class C (GTCC) Low-Level Radioactive
Waste and GTCC-Like Waste, DOE/EIS-0375, Washington, D.C. (2016).
[2] UNITED STATES NUCLEAR REGULATORY COMMISSION, Historical and
Current Issues Related to Disposal of Greater-Than-Class C Low-Level Radioactive
Waste, Enclosure 2. Technical Considerations Associated with Greater-Than-Class C
Low-Level Radioactive Waste Disposal and Qualitative Examination of Disposal
Challenges, SECY–15–0094, ADAMS Accession No. ML15162A821, Washington,
D.C. (2015).
[3] UNITED STATES DEPARTMENT OF ENERGY, Draft Environmental Impact
Statement for the Disposal of Greater-Than-Class C (GTCC) Low-Level Radioactive
Waste and GTCC-Like Waste, DOE/EIS-0375-D, Washington, D.C. (2011).
[4] INTERNATIONAL ATOMIC ENERGY AGENCY, Classification of Radioactive
Waste, IAEA Safety Standards Series No. GSG-1, IAEA, Vienna (2009).
[5] ORGANIZATION FOR ECONOMIC CO-OPERATION AND DEVELOPMENT \
NUCLEAR ENERGY AGENCY, Shallow Land Disposal of Radioactive Waste:
Reference Levels for the Acceptance of Long lived Radionuclides, A Report by an
NEA Expert Group, OECD, Paris (1987).
[6] INTERNATIONAL ATOMIC ENERGY AGENCY, Disposal Approaches for Long
Lived Low and Intermediate Level Radioactive Waste, IAEA Nuclear Energy Series
No. NW-T-1.20, IAEA, Vienna (2009).
[7] BRIMFIELD, T.C., et al., “Not All Greater-Than-Class C (GTCC) Waste Streams are
Created Equal,” Waste Management (WM2015 Conference, March 15-19, 2015,
Phoenix, Arizona, USA), Phoenix, AZ (2015).
[8] ARGONNE NATIONAL LABORATORY, Supplement to Greater-Than-Class C
(GTCC) Low-Level Radioactive Waste and GTCC-Like Waste Inventory Reports,
Washington D.C. (2010).
[9] UNITED STATES CONGRESS, An Evaluation of Options for Managing Greater-Than-
Class-C Low-Level Radioactive Waste,” OTA-BP-O-50, Office of Technology
Assessment, Washington, D.C. (1988).
[10] SANDIA NATIONAL LABORATORIES, Basis Inventory for Greater-Than-Class C
Low-Level Radioactive Waste Environmental Impact Statement Evaluations, Prepared
for U.S. Department of Energy, Washington, D.C. (1988).
Session 3c – ILW IAEA-CN-242
26
03c – 06 / ID 135. Disposal of Intermediate Level Waste
IMPLEMENTATION OF A GRADED APPROACH IN RADIOACTIVE WASTE
MANAGEMENT IN FRANCE
J.M. Hoorelbeke, S. Thabet
Andra, 1-7, rue Jean Monnet F-92298 Châtenay-Malabry cedex, France
E-mail contact of main author: [email protected]
Abstract. Andra is operating near-surface facilities at the industrial scale in France to dispose of very low
level and low level and short-lived wastes. Otherwise Andra’s deep geological Cigéo project is under
preparation to dispose of long-lived ILW and HLW, a large part of them resulting from spent fuel reprocessing.
In between those wastes that can be accommodated by near-surface existing facilities with respect to safety and
those wastes which require the high degree of isolation and containment provided by deep geological disposal, a
wide range of wastes will have to be managed in appropriate disposal facilities to be developed. Some are
legacy while others will be generated in the future. They include for instance radium bearing and other potential
NORM wastes as well as particular decommissioning radioactive waste such as graphite waste (recognized as
“low level long-lived waste” in France). Furthermore the diversity of decommissioning VLLW streams may
suggest dedicated disposal routes in the future.
IAEA’s Specific Safety Requirements SSR-5 provides that the ability of the chosen disposal system for a waste
type to provide its containment and to isolate it from people and the environment is to be commensurate with the
hazard potential of this waste in accordance with a graded approach. Hazards vary widely due to the diversity of
radioactive emission types and energies, of half-lives of radionuclides, of chemical properties and of bio-
toxicity. The needs for isolation and containment are to be formulated in terms of performance, for instance
retardation and mitigation, as well as in terms of a suitable assessment timescale. This timescale is to be defined
consistently with the potential reduction of activity of the waste with time and the evolution of the disposal site
and the containment system.
The definition of an appropriate disposal system includes the natural and/or engineered containment barriers, the
disposal depth, the site characteristics and their evolution over the considered timescale with regard to local
geodynamic conditions, the specific measures that may be implemented during the institutional control period
etc. Social acceptability is a crucial factor in determining proportionate solutions as well as siting disposal
facilities.
Within the framework of the French National Plan for the Management of Radioactive Materials and Waste,
Andra is developing a graded approach to propose new disposal options to complement the existing facilities
and the Cigéo project, in a view to optimizing the use of disposal capacities.
1. Taking into account the diversity of waste for a proper management
Today very low level and low level and short-lived radioactive wastes (VLLW, LLW) are
being disposed of in France by Andra in dedicated near-surface facilities. Otherwise high
level waste (HLW) and long-lived intermediate level waste (ILW) are planned to be disposed
of in deep geological Cigéo project under preparation.
A wide range of wastes may be considered as “in between”: their harmfulness makes them
unsuitable for surface disposal but does not necessary require geological disposal at great
depth. They include graphite waste, waste containing radium and some other waste such as
bituminized sludge from the treatment of effluents in nuclear facilities or maintenance
waste [1]. Most graphite waste comes from the dismantling of former natural uranium gas-
cooled reactors. Radium-bearing waste and broader NORM waste is mostly produced by non-
nuclear industrial activities.
Session 3c – ILW IAEA-CN-242
27
An example of the consideration of “in between” wastes is given by the work carried out with
the IAEA from 2013 to 2016 to discuss specifically the disposal of ILW and provide a
reference for selecting appropriate disposal concepts including a suitable depth. Indeed ILW
can be considered as in between LLW that are suitable for near-surface disposal and HLW
that require deep geological disposal. Similar principles may be applied to the broader work
to be carried in France to implement waste management solutions that aim at being
proportioned to the harmfulness of the wastes, consistently with IAEA SSR-5, which
provides that: “In accordance with the graded approach, as required in the International
Basic Safety Standards and other standards, the ability of the chosen disposal system to
provide containment of the waste and to isolate it from people and the environment will be
commensurate with the hazard potential of the waste” [2].
At the lower end of the range of radioactive wastes, most VLLW come from the dismantling.
Their activity level may widely vary and large volumes will arise in the future. A significant
part would be below the clearance level used in a number of other countries but not
considered in France: the disposal of this “very” very low level waste in dedicated facilities
may help to maintain traceability for future generations. Within this framework, the amount
and diversity of VLLW suggest adapting disposal solutions to the specificities of various
VLLW streams.
2. Harmfulness of wastes and needs for isolation and containment
Radioactive waste presents a potential hazard to human health and the environment and it
must be managed so as to ensure any associated risks do not exceed acceptable levels in the
short term as well as in the long term. As pointed out during IAEA technical meetings on the
safe disposal of ILW, hazards vary widely due to varying types of radioactive emissions,
varying energies of these emissions, half-lives of the nuclides in the waste as well as
chemical properties of various contented substances. In addition to the radiological hazard,
waste may also contain chemically toxic components, such as heavy metals. Contaminated
asbestos may also be present in nuclear facilities. Some radionuclides such as uranium
present both a radiological and a chemo-toxic hazard.
Waste management includes a number of successive steps such as sorting, treatment,
recycling as possible, conditioning and storage. End waste is to be disposed of. According to
IAEA safety standards, containment and isolation are the basic principles underpinning safe
disposal of waste to protect man and environment. The choice of a disposal solution needs to
ensure that these principles are met to the degree necessary for the waste during operation
and after closure of the facility. This degree of containment and isolation includes level of
performance as a function of time, taking into account the half-lives, activities and types of
the radionuclides in the waste to be disposed of.
Containment consists in preventing or controlling the release of radioactive substances and
their dispersion in the environment. Isolation is defined in SSR-5 as retaining the waste and
keeping its associated hazard away from the biosphere in a disposal environment that
provides substantial physical separation from the biosphere, making human access to the
waste difficult without special technical capabilities, and restricts the mobility of most of the
long lived radionuclides.
The radiological content of the waste, in terms of half-lives of predominant radionuclides and
in terms of activity level, is crucial to determine the time-scale required for containment and
isolation. Short lived radionuclides are usually considered with a half-life less than around
thirty years. The radiological harmfulness of waste with predominant short-lived
radionuclides significantly decreases within the time scale generally considered for the
Session 3c – ILW IAEA-CN-242
28
institutional control (a few hundred years). When considering long-lived radionuclides
defined by half-lives higher than thirty years, a particular attention should be given to the
diversity of these half-lives. Carbon 14, Radium 226 or Americium 241 are predominant in a
wide range of “in between” wastes. Provided the content of these wastes in very long-lived
radionuclides such as Chlorine 36 or Iodine 129 is limited, the required time-frame for
isolation and containment is of the order of 10,000 years. Such a time-frame may be
considered to be long with regard to human civilization, but it is moderate with regard to
geological evolutions, in particular in areas with low geodynamic processes. A much longer
time frame, 100,000 years and more, is required for HLW or spent fuel if considered as
waste:
The required level of performance in containment and isolation of radionuclides is a function
of the type and energy of emission, the activity level in the waste and the mobility of
elements in the geosphere and the biosphere. Regarding isolation in particular, these waste
specific characteristics determine the radiological impact in inadvertent intrusion scenarios as
a function of relating exposure routes (ingestion, external exposure). The needs for isolation
and containment are also to be adapted to the chemo-toxic harmfulness of the waste. Existing
regulation in non-nuclear fields may be used. Radiological and non-radiological harmfulness
should be managed consistently, both in terms of characterization of potential effects on
health and environment and in terms of their consequences on the needs for isolation and
containment. Assessing such a consistency probably requires significant work in the future,
especially when addressing low exposure levels, long time-scales as well as the consistency
with health and environmental protection in non-nuclear activities.
3. Development of proportioned disposal solutions
The first priority before defining disposal solutions is to reduce the volume and the
harmfulness of waste during production process as possible. In France reuse or recycling of
material is also recommended by the French environmental law for any waste – including
non-radioactive and radioactive – as the second priority (“Code de l’environnement”). And
finally end waste is to be properly disposed of. Sorting and treatment can be implemented to
help to reduce volume and/or harmfulness and therefore to facilitate their disposal. Fig. 1
illustrates the main components of waste management as provided by the French National
Plan for the Management of Radioactive Materials and Waste, issued every three years under
the auspices of the Ministry in charge of ecology and the regulator “Autorité de sûreté
nucléaire” [3].
FIG. 1. A typical waste management route according to the French National Plan for the
Management of Radioactive Materials and Waste [3]
Containment and isolation is provided by a combination of natural and engineered
characteristics of the disposal system. This issue has been addressed in detail during IAEA
technical meetings on the safe disposal of ILW, and can be applied to any type of waste:
Containment is achieved by maintaining package integrity, limiting the solubility of
radionuclides and the waste form, minimizing where possible groundwater inflow and/or
providing a long travel time for radionuclide transport from the disposal facility to the
Sorting Treatment Packaging Storage Disposal
Potential transport
Production
Session 3c – ILW IAEA-CN-242
29
biosphere; isolation is generally provided by depth of the disposal facility and to some extent
by the geology and environment surrounding the site.
In particular the selected depth of any disposal facility contributes to define the degree and
duration of isolation and of protection from surface erosion due to effects such as glaciation.
Nearer to the surface, natural changes occur over shorter timescales than deeper underground.
Significant processes leading to this evolution include erosion by wind, rain water,
weathering, climate induced processes such as glaciation, etc. These phenomena may change
the future boundary conditions of the system, for example, the hydrographic system and
hydrogeology, as well as the system itself, for example, through the changing chemical,
hydrological and temperature conditions. They will possibly progressively reduce the
thickness and/or performance of containment barriers interposed between the waste and the
environment. In an extreme situation the disposal facility and waste packages may be
destroyed in the long term, leading to loss of containment, direct access to waste and
dispersion of residual activity. The affected depth with time and the speed and consequence
of these mechanisms are site dependent.
The potential contribution to isolation of the institutional control and the memory keeping is
also important, where various time-scales may be considered.
4. Work in progress in France
Andra aims at developing a graded approach to propose new disposal options to complement
the existing facilities and the Cigéo project in connection with suitable predisposal
management options. This approach aims at an overall optimization of the use of disposal
facilities and of the distribution of wastes between these facilities with respect to safety and
cost. It will make it possible to manage all existing and future wastes in a consistent manner,
making the best use of available resources and avoiding undue burden on future generations.
This work requires strong interactions with various stakeholders, including the public. It is
part of the comprehensive approach offered by the French National Plan for the Management
of Radioactive Materials and Waste.
REFERENCES
[1] Inventaire national des matières et déchets radioactifs – www.andra.fr
[2] INTERNATIONAL ATOMIC ENERGY AGENCY, Disposal of Radioactive Waste,
Specific Safety Requirements No. SSR-5, IAEA, Vienna (2011).
[3] French National Plan for the Management of Radioactive Materials and Waste
(PNGMDR) - www.developpement-durable.gouv.fr .
Session 3c – ILW IAEA-CN-242
03c – 07 / ID 80. Disposal of Intermediate Level Waste
METHODOLOGY AND RESULTS FOR THE SAFETY ASSESSMENT FOR LOW-
AND INTERMEDIATE LEVEL WASTE REPOSITORY (SFR) IN SWEDEN
K. Källström1, E. Andersson
1, M. Lindgren
2, M. Odén
1, U. Kautsky
1, F. Vahlund
1,
J. Brandefelt1, P. Saetre
1, H. von Schenck
1, P.G. Åstrand
3, P.A. Ekström
3
1 Swedish Nuclear Fuel and Waste Management Company (SKB), Stockholm, Sweden
2Kemakta Konsult AB, Stockholm, Sweden
3Facilia AB, Stockholm, Sweden
E-mail contact of main author: [email protected]
Abstract. The Swedish low- and intermediate level waste repository, SFR, has been operating since 1988.
When the nuclear power plants in Sweden will be decommissioned and dismantled additional repository
capacity is required. Additional disposal capacity is also needed for operational waste from nuclear power units
in operation since their operating life-times have been extended compared with what was originally planned.
In December 2014, the Swedish Nuclear Fuel and Waste Management Company (SKB) submitted an
application to the Swedish Radiation Safety Authority (SSM) to extend the existing repository for low- and
intermediate level waste (SFR). SFR, the existing part and planned extension, is placed below the sea floor at
60-120 meter depth in Paleoproterozoic metagranite. For the application an evaluation of post-closure safety is
required. This paper presents the safety assessment performed to evaluate if the repository complies with the
Swedish Radiation Safety Authority’s regulations concerning safety and protection of human health and the
environment in the post-closure perspective. The results from the safety assessment are compared against the
annual risk criterion specified in the regulations, 10-6
, which corresponds to 1 % of the background radiation at
the site. The time frame of the safety assessment is 100,000 years under which there is an evolution of both the
repository and the external conditions (climate and surface systems).
The extended SFR repository and the applied 10 step methodology for the safety assessment are described.
Some steps of the methodology are discussed in more detail, e.g. the FEP-analysis and safety functions.
Different scenarios that will contribute to the overall risk evaluation for the repository are generated from
uncertainties in both external and internal processes. The understanding of the processes is based on extensive
site investigations, research, and numerical modelling of the evolution of the repository and external conditions.
Examples of major results for the dominating radionuclides (C-14, Mo-93 and Ni-59) are presented. The central
conclusion of the safety assessment is that the extended SFR repository meets the regulatory criterion and is
robust and safe in the post-closure perspective.
Key Words: Safety assessment, methodology, SFR
1. Introduction
This paper describes the methodology applied and results from the post-closure safety
assessment preformed to show compliance with the Swedish regulations [1] as a part of the
application to extend the existing Swedish repository for low- and intermediate level waste
(called SFR) situated in Forsmark. The extended SFR repository and the applied 10 step
methodology for the safety assessment are described.
Session 3c – ILW IAEA-CN-242
31
2. Description of the repository and the waste
SFR, the existing part and planned extension, is located at 60-120 meter depth in
Paleoproterozoic metagranite below the Baltic Sea floor. Due to land-rise after the last
glaciation, SFR and the overlying rock will in the future be situated below land instead of the
Baltic Sea. The waste is emplaced in different waste vaults consisting of engineered barriers,
adapted to the different protection needs of the particular waste forms. The purpose of the
barriers is to contain the radionuclides, and to prevent or retard the dispersion of those
substances, either directly or indirectly by protecting other barriers in the barrier system.
SFR, with its existing part and the planned extension contains one silo and 10 other waste
vaults (see FIG.1.). The design has been adapted to the properties of the wastes deposited in
each vault. The silo which contains the majority of the activity has both concrete and bentonite
barriers. The two waste vaults 1BMA and 2BMA consist of concrete structures, a waste vault
for boiling water reactor pressure vessels (BRT) is filled with grout, and in two waste vaults
(1BTF and 2BTF) the spacing between containers is filled with grout. For very low level
waste the only barrier is flow limiting plugs installed at closure. For a detailed description of
the repository and the waste vaults see [1].
According to the Swedish regulations the safety assessment for this type of repository needs
to cover a time period of at least 10,000 years but the required time frame is at most 100,000
years. On that time scale the engineered barriers of SFR will, to different degree, degrade.
There are two overall safety principles for SFR – limitation of the activity of long-lived
radionuclides and retention of radionuclides, see Section 2.1.2 in [2]. The activity of the
waste decreases with time, thus relaxing the demands on the protective capacity of the
degrading barriers over time.
3. Safety assessment methodology
The assessment methodology has been further developed since the most recent safety
assessment for SFR, SAR 08 [3], and is largely consistent with the methodology applied in
the safety assessment of the repository for spent fuel, SR-Site [4].The methodology applied
for the post-closure safety assessment SR-PSU consists of 10 main steps illustrated in FIG. 2.
FIG. 1. The existing SFR (light grey) and the extension (blue) with access tunnels. Illustration
reproduced from [2].
1.1 Step 1: Handling of FEPs
This step consists of identifying all factors that need to be considered in order to gain a good
understanding of the evolution and safety of the repository. This is done in a screening of
Session 3c – ILW IAEA-CN-242
32
potentially important features, events and processes (FEPs). Experience gained from previous
safety assessments of SFR, including SAR-08 [3], and international databases of relevant
FEPs that affect post-closure safety e.g. the NEA FEP-database [5] are utilised.
1.2 Step 2: Initial state
A thorough description of the waste, the repository and its environs at the time of closure is
needed as a starting point for all further evaluation of the post-closure safety of the
repository.
FIG 2. Overview of the ten steps in the methodology used for the post-closure safety assessment SR-
PSU. Illustration reproduced from [2].
1.3 Step 5: Definition of safety functions
This is a central step and consists of identifying and describing the repository system’s safety
functions and how they can be evaluated with the aid of a set of safety function indicators
that consist of measurable or calculable properties of the wastes, engineered barriers,
geosphere and surface system. The overall safety principles are broken down and described in
terms of a number of specified safety functions and safety function indicators. The fact that a
safety function deviates from its expected status does not necessarily mean that the repository
does not comply with regulatory requirements, but rather that more in-depth analyses are
needed to evaluate safety.
1.4 Step 6: Reference evolution
The purpose of the reference evolution is to provide an understanding of the overall future
evolution of the repository system including the uncertainties of importance for the post-
closure safety of the repository. The reference evolution is an important basis for the
definition of a main scenario and less probable scenarios. The reference evolution covers the
entire time period with an emphasis on the initial 1000 years which is required in Swedish
regulations [6]. The remaining time period consist of temperate climate conditions and
periglacial climate conditions.
Session 3c – ILW IAEA-CN-242
33
1.5 Step 8 and 9: Scenarios and calculation of radionuclide transport and dose
With the aid of the safety functions and the description of the reference evolution, a number
of scenarios are chosen to cover possible future evolutions of the repository system. A main
scenario and a number of less probable scenarios are analysed to examine whether the total
risk from all scenarios is below 10−6
. Doses are calculated both deterministically and
probabilistically in coupled box models that include the repository, the rock and the surface
system.
4. Radiological risk
The estimated radiological risks for the main scenario and each of the less probable scenarios
are presented in FIG. 3, The highest radiological risk is generally obtained for the main
scenario, except for a short period around 3000 AD when the highest risk is obtained by
human intrusion in the 1BLA waste vault. Mo-93, C-14, and Ni-59 contribute most to the
total radiological risk but at different time periods. For the entire time period the risk is below
the 10−6 regulatory risk criterion.
FIG. 3. Left: Radiological risk for each scenario taking into account the scenario-specific
probabilities. Right: Contribution to total radiological risk from each radionuclide. Illustrations
reproduced from [2].
5. Conclusions
Due to the combination of sufficiently limited activity of long-lived radionuclides and
sufficient retention of radionuclides in the repository, the central conclusion of the safety
assessment SR-PSU is that the extended SFR repository meets regulatory criteria on post-
closure safety.
REFERENCES
[1] The Swedish Nuclear Fuel and Waste Management Company, Initial state report for the
safety assessment SR-PSU. SKB TR-14-02, Stockholm (2014).
[2] The Swedish Nuclear Fuel and Waste Management Company, Safety analysis for SFR.
Long-term safety. Main report for the safety assessment SR-PSU. SKB TR-14-01,
Stockholm (2014).
[3] The Swedish Nuclear Fuel and Waste Management Company, Safety analysis SFR 1.
Long-term safety. SKB R-08-130, Stockholm (2008).
[4] The Swedish Nuclear Fuel and Waste Management Company, Long-term safety for the
final repository for spent nuclear fuel at Forsmark. Main report of the SR-Site project.
SKB TR-11-01, Stockholm (2011).
Session 3c – ILW IAEA-CN-242
34
[5] NEA, 2006. Electronic version 2.1 of the NEA FEP database developed on behalf of the
Nuclear Energy Agency by Safety Assessment Management Ltd.
[6] SSMFS 2008:37. The Swedish Radiation Safety Authority’s regulations concerning the
protection of human health and the environment in connection with the final management
of spent nuclear fuel and nuclear waste. Stockholm: Strålsäkerhetsmyndigheten (Swedish
Radiation Safety Authority).
Session 3c – ILW IAEA-CN-242
35
03c – 08 / ID 95. Disposal of Intermediate Level Waste
IMPLEMENTATION OF REQUIREMENTS ON THE CHEMICAL TOXICITY OF
NUCLEAR WASTE AT A REPOSITORY
A. Glindkamp1, B. Peschel
1, I. Harms
2
1TÜV NORD EnSys GmbH & Co. KG, Hanover, Germany
2Lower Saxony Water Management, Coastal Defence and Nature Conservation Agency
(NLWKN), Hildesheim, Germany
E-mail contact of main author: [email protected]
Abstract. In this contribution we will focus on non-radioactive harmful substances in a deep geological
repository. The implementation of specific requirements for the protection of groundwater against pollution is
exemplified by the repository Konrad. We will show how the protection target “protection of water against
pollution” is achieved.
The possible releases of non-radioactive harmful substances via water path were investigated within the scope of
the long-term safety assessment for the repository. Based on this investigation the license for the Konrad
repository was issued including the specific Water Law Permit, which handles the requirements concerning the
possible pollution of groundwater. The specifications of the permit were implemented by the operator of the
repository resulting in an adoption of the waste acceptance requirements.
Key Words: chemical toxicity, repository, groundwater
1. Introduction
In radioactive waste disposal radiological impacts as well as impacts of chemotoxic
components of radioactive waste packages must always be taken into consideration. The
radiological protection target “protection against ionizing radiation” and the protection target
of the near-surface groundwater “protection of water against pollution” have to be
considered.
The repository Konrad is a deep geological repository for radioactive waste with negligible
heat generation (low and intermediate level active waste). The repository is constructed in a
depth of 850 m within an iron ore formation of sedimentary origin, which reveals low, but
existent hydraulic permeability. 400 m of clayey strata above the repository are assumed to
be impermeable and thus form a hydraulic barrier.
2. Long Term Safety Assessment
The safety assessment to evaluate the influences of chemotoxic substances was made by the
Federal Office of Radiation Protection (BfS). It was presented in the IAEA-TECDOC-1325
[1]. The long term safety assessment was based on the scenario that the radioactive waste
with its non-radioactive harmful substances is assumed to come into contact with water
originating from the surrounding rock (‘formation water’) in the post-operational phase and
that non-radioactive harmful substances could be transported into the near-surface
groundwater. To minimize the calculation effort a conservative freshwater model was
applied.
In the meantime the license for the Konrad repository was issued including the specific Water
Law Permit, which is based on the long term safety assessment. The amount of non-
radioactive harmful substance is limited by the Water Law Permit. For 94 substances (e.g.
Session 3c – ILW IAEA-CN-242
36
lead, cadmium, toluol) a maximal disposable mass is determined. Further chemotoxic
substances may only be disposed in traces. This means that the quantity of these chemotoxic
substances in the repository is so low, that compromising the near-surface groundwater is
excluded. As it is laid down in the Specific Safety Requirements [2] the associated impact
indicators are given by water specific regulations. In the Water Law Permit it is also
determined that the composition of the deposited radioactive waste has to be monitored.
3. Implementation of the Water Law Permit
To meet these requirements BfS as operator of the repository Konrad has developed a
concept for monitoring the amount of non-radioactive harmful substances contained in the
radioactive waste packages. As one part of the concept, values for the content of non-
radioactive harmful substances in the waste packages (so called declaration threshold values)
are deduced to guarantee that the near-surface groundwater will not be affected.
The declaration threshold values are deduced for each substance, which shall be disposed of
in the repository. In this calculation the relevant limit values in the water specific regulations
are considered as well as substance-specific properties like solubility, composition or
estimated occurrence in the radioactive waste packages. Likewise, it is factored in that
different substances may exhaust the same limit values. For example, iron metal and readily
soluble iron salts both add to the exhaustion of limit value for dissolved iron. Thus, the sum
of the affection of these substances (plus further iron containing substances) on the near-
surface groundwater has to meet this value. The considerably lower solubility of iron metal
compared to readily soluble iron salts leads to a much higher declaration threshold value of
iron metal.
Harmful substances, which are enclosed in a mass fraction below their threshold values, are
classified as trace impurities and can be disposed of without balancing of their amounts. Only
those 94 harmful substances listed in the Water Law Permit can be disposed of in amounts
above their declaration threshold value. A so called material list is generated, in which an
entry for each substance that contains the threshold value and other specifications is
tabulated.
To simplify the description of radioactive waste packages, material vectors can be generated,
which are composed of other entries of the material list. Thereby different waste streams (e.g.
evaporator concentrates, ion-exchange resins) can be described easily by using a material
vector which was generated for this waste stream. In order to describe slightly different waste
streams, variations of the material vectors can be applied for at BfS. For material vectors the
declaration threshold values are deduced on the basis of the threshold values of the contained
substances and their portion in the material vector. Parallel to the material list a container list
is established, in which different containers are described on the basis of their materials.
The responsible water law regulatory authority Lower Saxony Water Management, Coastal
Defence and Nature Conservation Agency (NLWKN) with support of TÜV NORD EnSys
GmbH & Co. KG (TÜV NORD EnSys) as an independent expert organization has evaluated
this concept. It was determined that the concept is suitable to achieve the protection target of
the near-surface groundwater “protection of water against pollution”. Hence the NLWKN
agreed to the concept in 2011.
As a result of the above mentioned Water Law Permit implementation concept, the waste
acceptance criteria for the Konrad repository were adopted. It is now stated that the material
composition of all radioactive waste packages has to be described by the waste owner.
Packages which contain harmful substances above their declaration threshold value can only
Session 3c – ILW IAEA-CN-242
37
be disposed of if the contained substance is one of the 94 substances listed in the Water Law
Permit and the amount that is specified there is not exhausted yet.
4. Conclusion
The effect of the non-radioactive harmful substances in the repository Konrad was explored
by the investigation of possible releases via the water pathway in the post-operational phase
of the repository. To ensure that the amount of the non-radioactive harmful substances in the
repository Konrad is low enough to exclude a negative impact on the near-surface
groundwater, so called declaration threshold values were deduced. The evaluation by TÜV
NORD EnSys led to the conclusion that by this approach the protection target “protection of
water against pollution” can be certainly achieved.
By adopting the waste acceptance criteria the waste owners are committed to describe the
material composition of their waste packages. This description can be simplified by using
material vectors and container list entries. Due to the characterized composition of the waste
packages, BfS is able to monitor the materials, which are disposed of in the repository
Konrad, according to the requirements of the Water Law Permit.
The protection of groundwater is an important aspect concerning the disposal of radioactive
waste. The environmental impact of non-radioactive harmful substances should therefore be
investigated taking national regulations for the protection of groundwater into account.
REFERENCES
[1] IAEA-TECDOC-1325, Management of low and intermediate level radioactive wastes
with regard to their chemical toxicity, 2002
[2] IAEA Safety Standards Series No. SSR-5, Disposal of Radioactive Waste, 2011
Session 3c – ILW IAEA-CN-242
38
03c – 09 / ID 102. Disposal of Intermediate Level Waste
SAFETY ASSESSMENT AS AN INSTRUMENT FOR WASTE ACCEPTANCE
CRITERIA DERIVATION
A. Talitskaya1, E. Nikitin
1, A.Guskov
2, M. Nepeipivo
1, Sh. Garatuev
1, M. Rezchikov
1
1Scientific and Engineering Center for Nuclear and Radioactive Safety (SEC NRS), Moscow,
Russian Federation 2International Atomic Energy Agency (IAEA), Vienna, Austria
E-mail contact of main authors:
[email protected]; [email protected]; [email protected]
Abstract. According to requirements of Russian Federation regulatory framework the substantiation of safety
must be provided in the safety case report. One of the key parts of the safety case is the safety assessment. The
safety assessment must be performed at all stages of a facility lifecycle starting from facility siting and
development of conceptual design until the termination of the regulatory control usually linked to the period of
potential radioactive impact.
The safety assessment performed at designing stage of the near-surface disposal facility for operational and post-
closer period is presented here as a practical example. The main purpose of the safety assessment was a
derivation of maximum total activity and permissible specific activity for considered radionuclides in L/ILW.
Safety assessment for the operational period was performed according to the SADRWMS and GSG-3
methodologies for normal operation, accidental and incidental situations. Performed calculations resulted in the
doses that exceed the safety criteria for staff. Taking this into account permissible specific activity for
considered radionuclides were re-calculated as acceptance criteria.
Safety assessment for the post-closure period was performed according to the ISAM methodology. Normal
evolution scenario and alternative scenarios were considered. Obtained results exceed the admissible level of
radionuclide concentration in ground. Based on proportion of resulted concentration to allowable concentration
in ground the total permissible activity for each radionuclide was re-calculated.
After analysis of both operational and post-closure phases integrated waste acceptance criteria in terms of
radionuclide activity were derived for considered near surface disposal facility.
Key word: safety assessment, safety case, waste acceptance criteria, disposal
1. Introduction
Life cycle of disposal facility goes through several stages, including interrelated operation
and post-closure phases, and according to international practice it is assumed to distinguish
between long-term (post-closure) safety assessment (LSA) and operational safety assessment
(OSA). Operational and long-term safety assessments are widespread and admitted
instruments for objective analysis, assessment of possible radiation impact of radioactive
waste (RAW) disposal facility on human and the environment and decision making.
At the end of 1980th
Back End of the Nuclear Fuel Cycle became one of the most significant
problems of radiation safety for further nuclear energy development. LSA provides
understanding of a facility behavior over a long period. The main purpose of LSA is
estimation and analysis of radiological impact on human and environment due to
radionuclides migration from the RAW disposal taking into consideration wide range of
aspects – geological, chemical, physical, social and others. Our days widely used
methodology was developed within the IAEA Co-ordinated Research Project Improvement of
Session 3c – ILW IAEA-CN-242
39
Safety Assessment Methodologies for Near Surface Radioactive Waste Disposal Facilities
(ISAM) and then examined and illustrated within the Project on Application of Safety
Assessment Methodologies for Near Surface Radioactive Waste Disposal Facilities
(ASAM).Later on it was integrated into Safety Case within the following IAEA Projects:
Practical Illustration and Use of the Safety Case Concept in the Management of Near-
Surface Disposal (PRISM), Practical Illustration and Use of the Safety Case Concept in the
Management of Near-Surface Disposal Application (PRISMA). Result of these projects
became a base for further development of IAEA Safety Standards, such as SSR-5, SSG-23,
SSG-29 and etc. and regulatory documents in the Russian Federation NP-055-14, NP-058-14,
NP-069-14 and etc.
In comparison with long term timeframes of RAW potential hazard, the operational period
and operational safety previously considered as negligible. Only within the International
Intercomparizon and Harmonization Project On Demonstrating the Safety of Geological
Disposal (GEOSAF) it was realized that operational period can significantly affect the long
term safety of disposal facility. At the same time it was recognized in some countries that
safety of disposal facility during operation can’t be demonstrated just by the references to
radiation protection measures and emergency preparedness and response, but should be
somehow numerically assessed and ensured in a systematic manner. In general operation of
disposal facility is close enough to operation of storage facility and it seems to be possible to
use the methodology developed within the IAEA project on Safety Assessment Driving
Radioactive Waste Management Solutions (SADWRMS) and included into the IAEA General
Safety Guide No.3 “The Safety Case and Safety Assessment for the Predisposal Management
of Radioactive Waste” (GSG-3). Similar safety documents are under development in the
Russian Federation.
2. Practical example
For practical purposes one of real Near Surface Facilities for disposal of RAW of classes
3&41 was considered at design stage. The main purpose of the safety assessment was a
derivation of Waste Acceptance Criteria (WAC). Usually only long term (post-closure) safety
is considered for this purpose2 without taking into account operational period of disposal
facility. In this research both operational and long-term safety assessment were taken into
account.
Taken near surface disposal facility is a concrete vault with dimensions (length, width ,
height) - 150 × 25 × 7 m. Annual planned capacity is 1100 m3 of RAW. The whole capacity
of the disposal facility is 22000 m3 according to design. The operational time is supposed to
be at least 20 years. It is planned to place solid conditioned RAW in special concrete NZC
containers. After placing containers in NSF, filling free space by clay powder is assumed to
be performed. The composition of waste radionuclides include: U-238, Cs-137, Sr-90, Co-60.
For preliminary calculations maximum values of specific activity of considered radionuclides
as for RAW of the third class3 according to Russian legislation were used as an input.
1 According to the Governmental Decree No1069…
2 DERIVATION OF ACTIVITY LIMITS FOR THE DISPOSAL OF RADIOACTIVE WASTE IN NEAR
SURFACE DISPOSAL FACILITIES. IAEA, VIENNA, 2003. IAEA-TECDOC-1380 3 10
10 Bk/kg for β-radionuclides, 10
9 Bk/kg for α-radionuclides, 10
8 Bk/kg for transuranic radionuclides
Session 3c – ILW IAEA-CN-242
40
2.1.Long-term safety assessment
LSA includes calculation of radiation exposure on the population and the environment caused
by the possible withdrawal of radionuclides from the waste packages and their migration
beyond the safety barriers of disposal facility into the environment after the closure.
Calculations were performed for the maximal period of RAW potential hazard.
As safety indicators the values of specific activities in ground water on the sanitary protection
zone border were chosen. The following assumptions were made: NSF to be constructed,
commissioned, operated and finally isolated in accordance with the design; security,
environmental monitoring and physical control are supposed to be provided during the period
of active institutional control (first 50 - 100 years after the closure); structural integrity of
disposal will be preserved; NSF territory can’t be used by people for living and farming work
during the period of passive control (next 300 years).Normal evolution scenario and
alternative scenarios were considered when performing LSA.
Normal evolution scenario assumes that radionuclides from the waste matrix migrate through
containers, clay backfill and concrete wall of disposal vault into the environment. It was
assumed that the concrete does not change its strength and filtration properties during first
100 years. After 300 years since vault construction, concrete permeability corresponds
approximately to the permeability of sand.
In the period from 100 to 300 years, migration of radionuclides through concrete is due to
convection and diffusion processes, and over 300 years, is determined primarily by
convection. Migration of radionuclides through clay backfill is defined by diffusion process.
After migration through the safety barriers radionuclides get into the unsaturated zone and
further, by filtering with precipitation in the ground aquifer.
The migration of radionuclides in the aquifer is due to convective transport, taking into
account the physico-chemical processes (adsorption, ion exchange, etc.) and molecular
diffusion and hydrodispersion, which will be the scattering factor. As the alternative
scenarios considered "raising the groundwater level". This scenario consider changes in the
hydrogeological conditions at the site through the placement of the disposal 300 years,
despite the fact that the groundwater level rises above the base of the disposal. Because of the
degradation of engineering barriers in the system barriers will be enhanced permeability
zones ("filtration box"). Conservatively assumed that 100% of radionuclides are in the liquid
phase and can migrate with the flow of groundwater to drain, as in the normal evolution
scenario.
On the basis of the developed conceptual and mathematical models calculations using
Ecolego software tool have been conducted.During the LSA uncertainty and sensitivity
analysis were also carried out.
2.2.Operational safety assessment
Main aims of OSA for pre-closure waste management were evaluating of hazards and
radioactive impact on workers, population and the environment.
An individual dose rate for worker equal to 20 m/Sv, and for population – 0,1 m/Sv ware
used as safety criteria. For the environment – air, water and ground concentration (for
accidents and incidents) were used as safety criterion. According to the facility design
following workers are involved into operation of near surface disposal facility during its
Session 3c – ILW IAEA-CN-242
41
operational period: hoistman, slinger, dosimetrist, controller. The NSF is operated in a shift-
operation mode two times per week. Based on climate statistic it was supposed that 20% of
working days have adverse weather conditions that is why works at these days will be
missed. Total amount of operation modes per year was supposed as 80, average numbers of
containers per one mode is 8. There are 3 configurations of radioactive waste into NZC
container: 100% of Co-60, 10% of Sr-90 + 90% of Cs-137 and 100% of U-238. The
container value is 1,5 m3, wall thickness is 10 cm of concrete. For calculation it was supposed
that at each position works one employee. Next step of OSA was development of normal
operation, incidents and accidents scenarios. NZC protection uptakes α- β- radiation, that was
a reason for Sr-90 and U-238 exclusion from further consideration in normal operation
scenarios. Radiation impact for normal operation is due to external γ radiation of Co-60 and
Cs-137. However, in incident and accident scenarios consideration α-β-radiation may have a
serious impact due to internal exposure. As most dangerous accident scenario was considered
NZC drop with waste release. For each scenarios were developed conceptual and
mathematical models. Based on these models were calculated doses for workers and
population. Dose calculation with consideration direct and scattered radiation.
Operational safety assessment included uncertainties analysis. Uncertainties of time of
procedures may have affection on workers doses during all operational period upto 225%,
uncertainties of workers location relatively to containers – upto 210% and with both
uncertainties – upto 315%.
3. SA results and WAC derivation
Preliminary endpoint results of LSA excess of the safety criteria. Particular, calculations
shows exceeding of specific activity in water on the sanitary protection zone border for
radionuclide U-238 (3.0 Bq/kg according to national requirements for drinking water) when
the initial value of the activity in RAW is 109 Bq/kg. For safe disposal initial specific activity
of U-238 in a container was recalculated for WAC development. After recalculation following
initial activity of radionuclides were obtained: U-238 – 3,0∙105 Bq/kg; Cs-137, Sr-90 and Co-
60 – 1010
Bq/kg (no additional limitation). Preliminary OSA endpoint results also exided
the safety criteria - maximum allowable dose for workers –but for other than in LSA
radionuclides. Dose for public satisfy the safety criteria for normal operation, incident and
accident situations. Specific activity for WAC development were re-calculated based on OSA
results for 3 RAW composition: Co-60 (100%) – 8,94∙107 Bq/kg;
Sr-90 (10%)+Cs-137(90%) – 6,53∙109
Bq/kg; U-238 – 109
Bq/kg (no additional limitation).
OSA and LSA have resulted to different activity restriction. Integrated consideration of Waste
Acceptance Criteria for both LSA and OSA together gives following results:
Co-60 (100%) – 8,94∙107 Bq/kg (based on OSA, no LSA additional limitation);
Sr-90 (10%)+Cs-137(90%) – 6,53∙109
Bq/kg (based on OSA, without LSA additional
limitation); U-238 3,0∙105 Bq/kg (based on LSA, no OSA additional limitation). The
research result shows that just operational either just long-term safety assessment separately
is insufficient for determining those WAC parameters as radionuclide waste composition and
there acceptable specific activities.
4. Conclusion
In general LSA and OSA have similar structure and algorithm. However, scenarios,
instruments, assumptions and models are different. The main impact on WAC from LSA
results is caused by such factors as radionuclides half-life, engineered and natural safety
barriers retardation properties and the migration characteristic of radionuclides. Long-lived
Session 3c – ILW IAEA-CN-242
42
alpha and beta emitting radionuclides, such as uranium and transuranic elements, C-14 and
Cl-36 have the most impact on safety in long periods. It should be noted that carbon and
chlorine are neutral migrants that is practically not adsorbed by engineering barriers materials
and host rocks.
In case OSA the following factors appeared to be crucial: RAW management system, ,
equipment, number of workers and their qualification, safety culture. Gamma-emitting
radionuclides play the most critical role when considering normal operation. Alpha and beta
emitting radionuclides mainly have no any negative impact during normal operation, while
their presence may have a significant radioactive impact in case incidents and accidents.
According to WAC derivation the results show necessity of both operational and
long-term safety assessment to be carried out on the integrated approach basis. This works
concerns just radionuclide waste composition and there acceptable specific activities WAC
parameters, but there is sharp difference in WAC derivation results with separate
consideration from OSA or LSA standpoint. However, that is just fewer part of parameters and
other parameters derivation needs further researches based on the integrated approach.
Moreover, an integrated approach seems to be essential for other tasks, such as: development
and justification of technical, technological and organizational solutions of disposal;
development and justification of limits and conditions of safe operation and closure of
disposal; development and support of measures aimed at improving the safety of workers, the
population and the environment; justification for changes in the design of disposal etc.
Session 3c – ILW IAEA-CN-242
43
03c – 10 / ID 128. Disposal of Intermediate Level Waste
KONRAD REPOSITORY – EVALUATION OF THE SAFETY REQUIREMENTS
ACCORDING TO THE STATE OF THE ART OF SCIENCE AND TECHNOLOGY
B. Samwer, H. Baumgarten
Federal Office for Radiation Protection
E-mail contact of main authors: [email protected]
1. Introduction – Konrad repository and its history
The Konrad mine, an abandoned iron ore mine located in the area of the city of Salzgitter
(Federal State of Lower Saxony, Germany) is currently being converted to a repository for
radioactive waste with negligible heat generation (Intermediate Level Waste - ILW and Low
Level Waste - LLW). The overall responsibility for the construction and operation of the
Konrad repository is with the Federal Office for Radiation Protection (BfS).
Two shafts were sunk from 1957 to 1962 and the extraction of iron ore started in 1960.
Because of its favourable geology (Figure 1), the mine was investigated for its suitability to
host a repository for LLW and ILW as early as in 1976, after iron ore production had stopped
as a result of non-profitability. The iron ore deposit located in a depth of 1,300 m to 800 m is
12 to 18 m thick. However, the natural barrier in the form of clay and marl layers lying above
the mine is vital; being up to 400 m thick, it seals the mine from groundwater. On account of
the clay and marl layers, Konrad is an exceptionally dry mine, compared with other iron ore
mines.
In 1982, the Konrad mine was proposed as a repository for LLW and ILW with negligible
heat generation. At the beginning of 2007, a definitive plan-approval decision (licence) was
granted for the construction and operation of the repository by the Lower Saxon Ministry for
the Environment (NMU). Thus, the Konrad repository is the first facility for radioactive
waste management in Germany, for which a nuclear plan-approval procedure was conducted
prior to taking it into operation. The Konrad repository is permitted to take up max. 303,000
m³ of radioactive waste with a total activity of β- and γ-emitters of 5.0 · 1018
Bq and α-
emitters of 1.5 · 1017
Bq.
The two shafts of the Konrad mine are about 1.5 km apart. Shaft Konrad 1 serves for
personnel and material transport. Shaft Konrad 2 will serve as emplacement shaft. The
underground situation of the Konrad repository below ground is displayed in Figure 2.
2. Safety analyses for the Konrad mine
Comprehensive safety analyses were made in the scope of the plan-approval procedure for
the Konrad repository. Five aspects of safety analysis were investigated: 1. “Normal
operation”, 2. “Accidents”, 3. “Thermal influence on the host rock”, 4. “Criticality” and, 5.
“Long-term safety”. All safety analyses were examined by experts on behalf of the NMU and
compliance with specifications is controlled also by the state mining authority.
Session 3c – ILW IAEA-CN-242
44
FIG 1: Geological profile of the region of the Konrad mine showing the iron ore body of a thickness
of 12 m to 18 m. The future repository at a depth of 800 m to 1,300 m is covered by thick clay layers
of up to 400 m.
FIG 2: Underground situation of the Konrad repository. Excavation of the waste galleries has been
completed. Extensive work has been done on the surface facilities of Shaft 1 and the building site
equipment at Shaft 2 was set up.
These safety analyses determine requirements for the technical systems and components, the
operating procedures and the waste packages to be disposed of. They are binding in order to
guarantee safe operation and to minimise possible consequences. Furthermore, it was
investigated in long-term safety analyses how the repository could develop after it has been
sealed and possible consequences were derived. The long-term development of the Konrad
repository was forecast with the help of geo-scientific methods. In model calculations, the
dispersion of radionuclides from the repository up into the groundwater near the surface was
examined and evaluated. The model calculations show that it would take radionuclides at
least 300,000 years to get into the groundwater near the surface. For the transport of long-
lived radionuclides with a higher retention level in the geosphere, the model calculations
show relevant concentrations only after several million years. The calculated maximum
radionuclide concentrations that may occur in the groundwater near the surface have been
taken as a basis for the determination of the radiation exposure in the biosphere. For an
infant, the effective dose calculated according to the provisions set out in the Radiation
Session 3c – ILW IAEA-CN-242
45
Protection Ordinance is max. 0.26 millisieverts per year (mSv/a); for an adult it is max. 0.06
mSv/a. It is thus lower than the value of 0.3 mSv/a, this value having been applied for
evaluation by the licensing authority. Altogether, the possible impact on the near-surface
groundwater through the release of radionuclides and other pollutants from the repository is
so low that no adverse effects to man and environment need to be feared.
In addition to the safety analysis of normal operation, accidents were analysed. That means,
events in the planned operating procedures which might lead to a release of radioactive
substances into the environment were identified and evaluated. Technical or human failure
and rock-mechanical causes can be the reason for such accidents. In that context, the NMU
stated that the Konrad repository was designed in a manner that is balanced from the safety
point of view. Precaution required according to the state of the art of science and technology
has been taken against damage.
3. Evaluation of the Safety Requirements according to the state of the art of science
and technology
According to the current state of knowledge, there is no information available that is
questioning the safety statements given in the application documents. Furthermore, from the
legal point of view there is no breakpoint for the construction and operation of the Konrad
mine as a repository for radioactive waste. However, the BfS as a responsible owner and
operator has still provided for an evaluation of the safety requirements according to the state
of the art of science and technology prior to the repository being taken into operation. Here
within, the BfS sets a good example to improve safety standards and attempt to contribute to
increase trust and confidence into radioactive waste disposal.
The evaluation of the safety requirements according to the state of the art of science and
technology of the Konrad repository was initiated in 2014 and was continued with an expert
workshop to involve professional audience and stakeholders in April 2016. In the framework
of the workshop, safety-related aspects were collected, discussed and prioritised in three
working groups. The results of the working groups were published on the website of the BfS
and are taken into account in the work of the BfS. Targeted information of professional
audience and stakeholders will be continued at workshops and the public will be informed
continuously about the progress of the work via the internet.
The planned procedure of the BfS includes a step-by-step approach: 1. “Identification of
required updates of the safety analyses” and 2. “Update of safety analyses as required”
(Figure 3). The BfS coordinates and controls the entire process. Preparatory work is ongoing
and the phase, “Identification of required updates of the safety analyses” will be initiated by a
tendering procedure. The BfS will award the contract and the contractor will extensively
assess the safety statements given in the application documents. These safety analyses will be
compared according to the state of the art of science and technology (delta analysis).
Depending on the results, further steps will be conducted. The phase “Update of safety
analyses as required” will be executed if the assessment of the current safety analyses show
deviations from the state of the art of science and technology.
The work of the contractor will be continuously monitored by external experts via scientific
monitoring. The external experts will be installed by the regulatory authority, the Federal
Office for the Regulation of Nuclear Waste Management (BfE). The experts will discuss the
contractor’s results on a regular basis and advise on the work, if needed and the BfS will
coordinate the cooperation of the participants.
Session 3c – ILW IAEA-CN-242
46
To ensure neutrality and to control quality of the evaluation of the safety requirements, the
contractor’s work will be reviewed comprehensively (Peer Review) at the end of each phase.
The contractor will adapt the work accordingly and the BfS will compile the final results of
the work as well as the outcome of the scientific monitoring and the Peer Review to prepare a
final judgment about the evaluation of the safety requirements according to the state of the art
of science and technology.
In case that evaluation of the safety requirements shows that technical adjustments are
required, the BfS will adjust the planning and adapt possible technical changes prior to taking
the Konrad repository into operation. A periodic evaluation of the safety requirements
according to the state of the art of science and technology will continue after the Konrad
repository has been taken into operation (Figure 3). The entire process will be documented
carefully to prepare guidelines for future safety assessments.
FIG3: Periodic evaluation of the safety requirements according to the state of the art of science and
technology. The planned procedure will be monitored continuously by external experts and the results
will be reviewed comprehensively (Peer Review).
4. Conclusion
The Konrad mine is the first repository in the Federal Republic of Germany which has been
and will be planned, constructed, operated and sealed pursuant to the stringent specifications
of nuclear law, from the beginning of filing the application until the sealing of the mine later
on.
There is no information available that is questioning the safety statements given in the
application documents at this point in time and an evaluation of the safety requirements is not
required by law prior to the repository being taken into operation. However, the BfS as a
Session 3c – ILW IAEA-CN-242
47
responsible owner and operator proceeds with the assessing of the safety requirements
according to the state of the art of science and technology even now.
Neutrality and transparency need to be ensured throughout the entire procedure and the BfS is
monitored by the federal state of Lower Saxony and by the BMUB.
Further on, additional scientific monitoring of external experts, installed by the regulator BfE
and Peer Review of the evaluation of the safety requirements according to the state of the art
of science and technology is used for the purposes of neutrality. Together with targeted
public relations work, transparency of the process will be promoted to improve public
acceptability of the repository.
Future periodic evaluation of the safety requirements will be ensured by preparing guidelines
based on the current approach. Furthermore, the BfS will continuously monitor the
development of the state of the art of science and technology.