SESSION 3c Disposal of Intermediate Level Waste

Transcription

1 International Conference on the Safety of Radioactive waste Management SESSION 3c Disposal of Intermediate Level Waste

2 ORAL PRESENTATIONS No. ID Presenter Title of Paper Page 03c R. Nakabayashi Japan Development of Methodology for Probabilistic Safety Assessment of Long Term Radioactive Waste Disposal 4 03c S. Konopaskova Czech Republic Waste Acceptance Criteria Development for Different Low and Intermediate Level Waste (LILW) Disposal Systems 8 03c E. Andersson Sweden Assessment of Human Intrusion and Future Human Actions Example from the Swedish Low and Intermediate Level Waste Repository SFR 13 03c A. Carter United Kingdom Data Management to Support a Post-Closure Safety Case for Higher Activity Wastes 17 03c H. Arlt United States of America Greater-Than-Class C Low Level Radioactive Waste Characteristics and Disposal Aspects 21 03c J.-M. Hoorelbeke France Implementation of a Graded Approach in Radioactive Waste Management in France 26 2

3 POSTER PRESENTATIONS No. ID Presenter Title of Paper Page 03c K. Källström Sweden Methodology and Results for the Safety Assessment for Low and Intermediate level Waste Repository (SFR) in Sweden 30 03c A. Glindkamp Germany Implementation of Requirements on the Chemical Toxicity of Nuclear Waste at a Repository 35 03c M. Nepeypivo (A. Talitskaya) Safety Assessment as an Instrument for Waste Acceptance Criteria Derivation 38 Russian Federation 03c B. Samwer Germany Konrad Repository Evaluation on the Safety Requirements according to the State of the Art of Science and Technology 43 3

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 contact of main author: r-naka@criepi.denken.or.jp 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. 4

5 2.2.Demonstration of compliance with a dose criterion It is possible to demonstrate the compliance with a dose criterion by comparing the 95 th 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 95 th percentile of the dose distribution incorporating the uncertainties involved is less than the dose criterion. By demonstrating that the 95 th 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 95 th 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 95 th 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 95 th 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 50 th 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 (50 th 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 (50 th 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]. 5

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 ( Bq) is instantaneously released from radioactive waste in this model 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 lognormal 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]. 6

7 FIG. 3. PDF and CDF indicating the maximum annual dose of 14 C [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 msv/year, whereas the 50 th percentile of the CDF was msv/year, i.e., the 50 th percentile was larger than the mode. In this case, the 50 th percentile is adopted as the assessment result for comparison with the dose criteria in the likely scenario. The 95 th 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 95 th 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). 7

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 contact of main author: konopaskova@surao.cz 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 Operated repositories and their types The overview of repositories is specified in Table 1. 8

9 TABLE I: MARGINS FOR YOUR MANUSCRIPT. Site Type Volume Waste streams Matrix Host rock Dukovany, Richard, Bratrství, subsurface m 3 Waste from NPPs Bitumen, geopolymer LILW m 3 Institutional waste, artificial radionuclides LILW m 3 Institutional waste, natural radionuclides Cement Cement Crystalline Limestone mine Uranium mine FIG. 1. Dukovany repository a subsurface vault system 2.2.WAC structure FIG. 2. Richard and Bratrství repositoris underground disposal systems 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 9

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. 10

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 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. 3. Conclusions FIG. 3.Limits of activity in vault and underground systems 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). 11

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 12

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. Andersson 1, T. Hjerpe 2, G. Smith 3, K. Källström 1, L. Morén 1, K. Skagius 1 1 Swedish Nuclear Fuel and Waste Management Co., Sweden 2 Facilia AB, Sweden, 3 GMS Abingdon Ltd, UK contact of main author: Eva.Andersson@skb.se 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 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 13

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 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 14

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 15

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 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 ," 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 (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 (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. 16

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 contact of main author: alexander.carter@nda.gov.uk 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. 17

18 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 18

19 (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. 19

20 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 [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

21 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 contact of main author: hans.arlt@nrc.gov Abstract. United States regulations (Part 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 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 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 21

22 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 m 3 (311,000 ft 3 ) and the projected activity of that waste by 2083 will be 5.92 x 10 6 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 m 3 (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]. 22

23 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 10 5 TBq (4.1 MCi). The projected volume of GTCC commercial sealed sources is 2,900 m 3 (102,000 ft 3 ) [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 10 4 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 m 3 (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 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 23

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. 24

25 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 , 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). 25

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 Châtenay-Malabry cedex, France contact of main author: jean-michel.hoorelbeke@andra.fr 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 biotoxicity. 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 gascooled reactors. Radium-bearing waste and broader NORM waste is mostly produced by nonnuclear industrial activities. 26

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 27

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]. Production Sorting Treatment Packaging Storage Disposal Potential transport 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 28

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 [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)

30 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. Källström 1, E. Andersson 1, M. Lindgren 2, M. Odén 1, U. Kautsky 1, F. Vahlund 1, J. Brandefelt 1, 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 2 Kemakta Konsult AB, Stockholm, Sweden 3 Facilia AB, Stockholm, Sweden contact of main author: klas.kallstrom@skb.se Abstract. The Swedish low- and intermediate level waste repository, SFR, has been operating since 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 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.

31 2. Description of the repository and the waste SFR, the existing part and planned extension, is located at 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 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 31

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 postclosure 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. 32

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 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 postclosure 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 , 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). 33

34 [5] NEA, 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). 34

35 03c 08 / ID 95. Disposal of Intermediate Level Waste IMPLEMENTATION OF REQUIREMENTS ON THE CHEMICAL TOXICITY OF NUCLEAR WASTE AT A REPOSITORY A. Glindkamp 1, B. Peschel 1, I. Harms 2 1 TÜV NORD EnSys GmbH & Co. KG, Hanover, Germany 2 Lower Saxony Water Management, Coastal Defence and Nature Conservation Agency (NLWKN), Hildesheim, Germany contact of main author: aglindkamp@tuev-nord.de 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 nonradioactive harmful substance is limited by the Water Law Permit. For 94 substances (e.g. 35

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 nonradioactive 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 nearsurface 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 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 36

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,

38 03c 09 / ID 102. Disposal of Intermediate Level Waste SAFETY ASSESSMENT AS AN INSTRUMENT FOR WASTE ACCEPTANCE CRITERIA DERIVATION A. Talitskaya 1, E. Nikitin 1, A.Guskov 2, M. Nepeipivo 1, Sh. Garatuev 1, M. Rezchikov 1 1 Scientific and Engineering Center for Nuclear and Radioactive Safety (SEC NRS), Moscow, Russian Federation 2 International Atomic Energy Agency (IAEA), Vienna, Austria contact of main authors: anna.talitskaya@gmail.com; e.m.nikitin@yandex.ru; avguskov@gmail.com 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 postcloser 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 1980 th 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 38

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 , NP , NP 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&4 1 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 purpose 2 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) m. Annual planned capacity is 1100 m 3 of RAW. The whole capacity of the disposal facility is m 3 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 class 3 according to Russian legislation were used as an input. 1 According to the Governmental Decree No DERIVATION OF ACTIVITY LIMITS FOR THE DISPOSAL OF RADIOACTIVE WASTE IN NEAR SURFACE DISPOSAL FACILITIES. IAEA, VIENNA, IAEA-TECDOC Bk/kg for β-radionuclides, 10 9 Bk/kg for α-radionuclides, 10 8 Bk/kg for transuranic radionuclides 39

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 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 40

41 operational period: hoistman, slinger, dosimetrist, controller. The NSF is operated in a shiftoperation 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 % of Cs-137 and 100% of U-238. The container value is 1,5 m 3, 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 10 9 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, Bq/kg; Cs-137, Sr-90 and Co 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, Bq/kg; Sr-90 (10%)+Cs-137(90%) 6, Bq/kg; U 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, Bq/kg (based on OSA, no LSA additional limitation); Sr-90 (10%)+Cs-137(90%) 6, Bq/kg (based on OSA, without LSA additional limitation); U-238 3, 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 41

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. 42

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 contact of main authors: bsamwer@bfs.de 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 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 Bq and α- emitters of 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 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. 43

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 longlived 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 44

45 Protection Ordinance is max millisieverts per year (msv/a); for an adult it is max 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 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. 45

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 46