Disposal of High Level Waste

Transcription

1 International Conference on the Safety of Radioactive waste Management SESSION 3d Disposal of High Level Waste Including Spent Nuclear Fuel Declared as Waste

2 ORAL PRESENTATIONS No. ID Presenter Title of Paper Page 03d 00 INV 03d J. Heinonen Finland Regulatory Experiences from the Spent Fuel Disposal Step-Wise Implementation 4 03d V. Havlová Czech Republic Complex Safety Assessment Model of Radioactive Waste Deep Geological Disposal in the Czech Republic 7 03d A. Hagros Finland Preparing Posiva s Post-Closure Safety Case Towards the Operational Phase 11 03d S. Voinis France Andra s Safety Options of French Underground Facility Cigeo a Milestone towards the Licensing Application 15 03d I. Niemeyer Germany Bridging Nuclear Safety, Security and Safeguards at Predisposal and Geological of High Level Waste and Spent Nuclear Fuel 20 03d T. Fujiyama Japan Development of the NUMO Pre-selection, Sitespecific Safety Case 25 03d L. Bailey United Kingdom Development of a Generic Environmental Safety case for the Disposal of Higher Activity Wastes in the UK 29 03d D. Pellegrini France SITEX, the European Network of Technical Expertise Organisation for Geological Disposal 33 03d A. Ström Sweden Research and Development Needs in a Step- Wise Process for the Nuclear Waste Programme in Sweden 38 2

3 Session 3c ILW POSTER PRESENTATIONS No. ID Presenter Title of Paper Page 03d F. Charlier Germany Germanys New Route towards a Repository for High Level Waste Scientific Challenges 43 03d J.-W. Kim Korea Recent Safety Assessment of a Reference Geological Disposal System for Radioactive Waste from Pyro-Processing in Korea 47 03d Y. Kovbasenko Ukraine Assessment of Decay Heat in Process of Spent Nuclear Fuel Disposal 51 03d S. Suzuki Japan Assessment of Pre- and Post Closure Safety in the NUMO Safety Case for a Geological Repository 56 03d J. Stastka Czech Republic Research, Development and Demonstration Projects at the Josef Underground Laboratory 60 03d V. Maree South Africa The Management of Used (Spent) Fuel and High Level Waste in South Africa 64 03d J. Leino Finland Regulatory Experiences in Reviewing Construction License Application for the Disposal of Spent Nuclear Fuel in Finland 69 03d L. Vondrovic Czech Republic Generic Underground Research Facility in the Middle Stage of the Site Selection Process: Bukov URF, Czech Republik 73 03d F. Launeau France Cigeo Project: from Basic Design to Detailed Design pursuant to Reversibility 77 03d B.B. Acar Turkey Impact of Storage Period on Safe Geological Disposal of Spent Fuel 81 3

4 03d 00 / INV 03b. Disposal of High Level Waste REGULATORY EXPERIENCES FROM THE SPENT FUEL DISPOSAL STEP-WISE 1. Introduction IMPLEMENTATION J. Heinonen Radiation and Nuclear Safety Authority (STUK), Finland contact of main author: jussi.heinonen@stuk.fi Finland began planning and preparing for nuclear waste management measures in the 1970s, during the procurement and construction phase of the first nuclear power plants. In 1983, the Finnish Government made a policy decision on the principles and schedules of nuclear waste management. In 2000, the Government adopted a favourable Decision-in-Principle (DiP) accepting the concept of a deep disposal facility for spent fuel from the Finnish nuclear power plants in Olkiluoto and Loviisa. This DiP was confirmed by the Parliament in A construction licence application (CLA) for the encapsulation and disposal facility was submitted to the Government in The Finnish government granted in 12th November 2015 Posiva license to construct spent nuclear fuel (SNF) encapsulation plant and disposal facility in Olkiluoto. Government decision was supported with STUK s safety evaluation. Before encapsulation and disposal process begins, Government has to issue a operating license. Operating license application is expected to be submitted in early 2020 s and disposal is planned to start STUK, as the independent safety regulator, has been performing stepwise review of developing safety case and R&D needed to demonstrate safety of SNF disposal. STUK strategy has also been that safety regulation is developed coincide with the development of disposal. This approach has enabled to include experiences and growing knowledge timely into safety regulation. STUK also decided to participate actively in pre-siting and pre-license phase. Active participation has included pro-active public communication and step-wise evaluation of site characterisation work and development of safety case. Based on our experiences active participation and communication with implementer has been one of the key factors in regulatory work to enable effective progress in disposal development and licensing. 2. Regulatory activities during the step-wise implementation The licensing procedure for a disposal facility has several steps that are similar to all nuclear facilities in Finland and are defined in Nuclear Energy Act and Degree. These licensing steps are Decision-in-Principle (DiP), Construction License and Operating License. The first licensing step towards a disposal facility for spent nuclear fuel was Decision-in- Principle (DiP). As part of this decision Government decided that SNF would be disposed in 4

5 Session 3c ILW Olkiluoto using KBS-3 concept. In addition to the permit to proceed with the project, DiP gave also Posiva the authorization to start to construct an underground rock characterization facility (URCF) at the proposed site, to the depth of actual planned disposal, as required by safety regulation. After the DiP, STUK started work aiming for the readiness to review the construction license. One of the major activities was the regulatory oversight of the construction work of the underground rock characterization facility (URCF), Onkalo. STUK planned and executed the regulatory oversight of the URCF in similar manner as it would do for nuclear facility due to the fact that Posiva s plan is to use the constructed URCF as part of the disposal facility in the future. Besides the oversight of the construction work of the URCF, STUK followed closely Posiva s R&D work based on the R&D program published every third year and reviewed the draft post closure safety case documentation published by Posiva before year The aim of the step-wise review, close follow-up and regular meetings with Posiva was to identify the safety relevant issues and especially key safety concerns already before Posiva finalizes and submits the construction license application. The review of safety case parts was not always usefull in solving safety relevant issues and from this experience a need for more structured review and assessment process for the construction license application review was seen necessary. In addition to the activities related to Posiva, STUK also developed it s own resources and competence to prepare itself for the construction license review. In 2006 STUK management made a strategic plan to increase waste management resources before Posiva submits the CLA. Plan was followed and the amount of people working mainly for the waste management regulations was almost tripled during next six years. STUK made also framework contracts with 13 external experts to support STUK during in the review of CLA. STUK s task in the CLA process is to review and assess the fulfillment of all applicable radiation and nuclear safety requirements and prepare statement and safety evaluation report for the Government. STUK submitted statement regarding Posiva s CLA to the Govenrment in February STUK s main conclusion was that the planned encapsulation plant and disposal facility can be built to be safe. Also there is sufficient reliability that there will be no detrimental radiation effects to the public or environment neither during the operational period nor after decommissioning and closure of the facility. In the statement to the government STUK raised areas that need further development before specific construction step or before submittal of operating license application. 3. Conclusions Based on the experiences of regulating Posiva s DGR development, we have concluded that following aspects are important for effective regulatory work: Development and maintaining of up-to-date requirements. Requirements can be develop along with DGR development as more information and knowledge are gathered 5

6 Development of oversight strategy for each different phase. Starting for early conceptualization and siting phase regulatory functions and focus can differentiate a lot. In early phase regulators review can be more generic and evaluating that safety requirements could be met. In licensing steps however regulator has to conclude if safety requirements are met or not. This is the most challenging part of review and assessment and therefore clear criteria should be developed. Active interaction with implementer is needed for mutual understanding. Regulator has a important role in communicating with public and this involvement should start in early phase of repository development 6

7 Session 3c ILW 03d 01 / ID 111. Disposal of High Level Waste COMPLEX SAFETY ASSESSMENT MODEL OF RADIOACTIVE WASTE DEEP GEOLOGICAL DISPOSAL IN THE CZECH REPUBLIC V. Havlová 1, D. Trpkošová 1, A. Vokál 2 1 ÚJV Řež, a.s., Husinec, Czech Republic 2 SURAO, Dlážděná 6, Praha, Czech Republic contact of main author: vaclava.havlova@ujv.cz Abstract. A complex safety assessment (SA) model employing GoldSim software has been under development in the Czech Republic since 2006 aimed at proving the long-term safety of the country s future deep geological repository (DGR) over a period of 1 million years. The input data for each of the components of the model has been compiled from archive sources, expert literature and supporting research. The main concern with respect to the model is to adhere as closely as possible to conditions which will prevail within the real DGR by means of performing either laboratory or in-situ research. The paper includes a description of the model and examples of supporting research concerning both the near- and far-fields. Key Words: safety assessment, deep geological repository, GoldSim, radiological impact 1. Introduction The main aim of deep geological repository (DGR) safety assessments (SA) is to consider the performance of the repository system in terms of radiological impact or other global indicators of the impact on safety. Such assessments may vary in terms of the relevant time frame(s), the level of detail, the range of issues considered, the degree of precision required with respect to the input data and the resulting calculations. The reason for the safety case as well as the programme development stage often dictate both the scope of and degree of detail required in the safety assessment [1]. Consequently, a complex SA model employing GoldSim software has been under development in the Czech Republic since 2006 the purpose of which is to illustrate the longterm safety of the future Czech deep geological repository (DGR) over a period of 1 million years. The input data required for SA modelling purposes, consisting of results obtained from both archive sources and limited own research, was collected in 1999 and Currently, with respect to the performance of SA supporting research, the main concern is to adhere as closely as possible to conditions which will prevail within the real DGR. Therefore, it is essential that the research includes both laboratory and in-situ experimentation. 2. Czech disposal concept The Czech deep geological repository (DGR) concept assumes that waste packages containing spent nuclear fuel (SNF) assemblies will be enclosed in steel-based canisters placed in vertical or horizontal boreholes at a depth of ~ 500m below the surface. The void between the canisters and the host crystalline rock will be backfilled with compacted bentonite which will make up the final engineered barrier. The reference SNF canister consists of two protective layers, an outer layer of carbon steel which will corrode very slowly under anaerobic conditions and a second inner layer of stainless steel which will corrode at an almost negligible general corrosion rate and exhibit a low tendency to local corrosion under anaerobic conditions. It is presumed that the buffer material will originate 7

8 from Czech Republic bentonite deposits; currently, so-called Rokle bentonite (Ca-Mg bentonite) is being employed for experimentation purposes. In addition to SNF and high-level waste, intermediate-level waste (ILW) containing longlived radionuclides such as decommissioned reactor core parts and serpentinite concrete which does not meet the criteria for disposal in near-surface repositories will also be disposed of in the future DGR. However, ILW will be disposed of in a separate section of the repository to that of SNF assemblies since it is essential that the potential for the SNF and ILW to exert an impact on each other be avoided. ILW will be emplaced in concrete canisters in specially excavated chambers that will then be back-filled with a bentonite-based material. 3. Near-field The near-field SA model assumes the disposal of a total of 5800 carbon-steel canisters containing spent fuel (SF) with a minimum canister life-time of 10,000 years and a median canister life-time of 110,000 years. It is assumed that the release of radionuclides will occur following the degradation of the canisters. SF canister degradation is simulated by means of a distribution curve obtained via the application of the Weibull distribution. The original version of the model assumed a uniform inventory, however the latest version enables the inventory to be divided according to a number of defined preferential transport directions, each characterised by an individual transport pathway towards the surface [2]. The data which characterises carbon-steel canister material corrosion rates was obtained from the results of previous projects involving a limited number of laboratory experiments [2]. Current research projects however include both laboratory and real rock massif scale experimentation. Carbon steel, titanium and copper corrosion on contact with Rokle bentonite has been extensively investigated at the UJV s laboratories under defined anaerobic conditions as part of a previous project [3] - see Fig. 1. Further research in this respect was conducted in the context of the international Material Corrosion Test (MACOTE) project performed at the Grimsel test site [5], as part of which in 2015 five heated modules (of UJV construction) containing corrosion samples (steel, copper; Czech Ca-Mg bentonite and MX- 80 bentonite) were inserted into the rock massif up to a depth of 5 meters (anaerobic conditions). The modules will be extracted over a defined time-line of 1, 2, 3, 5 and 7 years. The results of both projects will subsequently be combined. Bentonite_Cell1_440 Bentonite_Cell2_440 Bentonite_Cell3_440 Bentonite_Cell4_440 Bentonite_Cell5_440 Bentonite_Cell6_440 Bentonite_Cell7_440 Bentonite_Cell8_440 Bentonite_Cell9_440 Bentonite_Cell10_440 Fig. 1. Carbon-steel sample in contact with Ca- Mg bentonite [3]. Bentonite_Cell11_440 Bentonite_Cell12_440 Bentonite_Cell13_440 Bentonite_Cell14_440 Bentonite_Cell15_440 Fig. 2. Bentonite layer representation in the GoldSim model. It is currently supposed that Rokle bentonite (Ca-Mg bentonite) will be used as the buffer material surrounding the disposal canister. For modelling purposes, the rock diffusion layer is considered to be the bentonite/rock compartment interface thus eliminating the influence of advection within the bentonite layer. The bentonite buffer layer is modelled in the form of 15 concentric layers (see Fig. 2), the outer layer of which represents the interface with the rock 8

9 Session 3c ILW massif enclosing the repository (the rock compartment). Radionuclides are transported by means of diffusion through the bentonite layers towards the rock compartment. Subsequently, the radionuclides are transported by means of groundwater flow from the near-field boundary towards a preferential path within the geosphere. Radionuclide diffusion data for safety assessment purposes is usually obtained via the conducting of laboratory through-diffusion experiments using radioactive tracers. Throughdiffusion experiments are based on the diffusive transport of tracers through the bentonite in the direction of the concentration gradient. geosphere_shallow_pathway geosphere_midle_pathway depository_closed_area geosphere_deep_pathway FIG. 3. GoldSim geosphere model [2] FIG. 4. Fracture model in PAMIRE project -preliminary results [7] In the case of bentonite, a process description is important particularly with respect to anionic radionuclides (e.g. I, Se, Tc) where relative retardation is anticipated due to anionic exclusion. The phenomenon has been described as part of the research outlined in [2], [3] etc. 4. Far-field The rock massif is modelled in the form of a compartment with dimensions of 3km 1km 10m while the geosphere is modelled using Pipe components which consider advection, dispersion, diffusion into the rock matrix and sorption as the principal processes under way. Groundwater flows into the compartment that models various processes at work in the biosphere from the final Pipe (see Fig. 3). Radionuclide migration processes have been studied under both laboratory (e.g. [2], [3], [4]) and in-situ conditions (e.g. [7], [8]). Whilst laboratory results are able to provide results for well-defined conditions, they are not able to fully reflect the conditions of the rock massif. Supporting in-situ research has been conducted at for example the Josef Underground Research Laboratory (CZ) (e.g. the PAMIRE project [7]) and at the Grimsel test site (Long Term Diffusion, LTD project [8], [9]). The PAMIRE project described a rock fracture in detail in preparation for the conducting of a migration advective test with radionuclides in granitic rock (see Fig. 4), whereas the Long Term Diffusion experiment project Phase III focused principally on the matrix diffusion process involving the injection of a radioactive cocktail consisting of 3 H, 22 Na, 133 Ba, 134 Cs and non-active Se(VI) into a granitic rock massif in 2014 and the subsequent observation of tracer diffusion [9]. 5. Biosphere The biosphere is modelled using four compartments representing land (cultivatable and forest), a lake and a river. The model represents a universal system that corresponds to the current lifestyle of the Czech Republic. The output of the biosphere model consists of the effective dose rate received by one member of the critical group in the environment. 9

10 6. Conclusions The SA model was not designed as a static model, rather the aim is to continue the development of the model so as to eventually describe the site finally chosen for the Czech DGR. The following aspects should be considered in the near term: the source term, the refinement of the geosphere transport model, the construction of individual biosphere models for each DGR candidate site, uncertainty evaluations etc. Work to date will be concluded with an SA evaluation due to be completed in 2018 which will address in greater detail one of the potential sites for the construction of the Czech deep geological repository. 7. Acknowledgement The research presented in this study was funded by SURAO [2, 3, 5, 8], the Ministry of Trade and Industry [6] and the Czech Technology Agency (TAČR) [7]. REFERENCES [1] NEA-OECD (2012): Methods for the Safety Assessment of Geological Disposal Facilities for Radioactive Waste (MeSA). NEA No. 6923, OECD, [2] Scientific support of DGR safety assessment. SURAO project ( ). [3] Research and development of a disposal canister for SNF deep geological disposal. SÚRAO project ( ). [4] Research of material properties for the safe disposal of radioactive wastes and the development of procedures for their evaluation. MPO TIP FR TI-1/362 project. [5] [6] Research on the influence of inter-grain granite permeability for safe disposal in a geological formation; methodology and measurement device development; MPO TIP FR TI-1/367 [7] PAMIRE - TA [8] Long-term diffusion Phase VI. project - [9] Soler J. and Martin A. (2015): LTD Experiment Monopole 2: Predictive Modeling and Comparison with Initial Monitoring Data. NAGRA Report NAGRA, Wettingen, Switzerland. 10

11 Session 3c ILW 03d 02 / ID 130. Disposal of High Level Waste PREPARING POSIVA S POST-CLOSURE SAFETY CASE TOWARDS THE OPERATIONAL PHASE A. Hagros 1, H. Reijonen 1, B. Pastina 2, N. Marcos 1, P. Hellä 1 1 Saanio & Riekkola Oy, Helsinki, Finland 2 Posiva Oy, Eurajoki, Finland contact of main author: annika.hagros@sroy.fi Abstract. Posiva Oy is currently preparing a safety case to support the operating licence application (OLA) for the spent nuclear fuel disposal facility under construction at the Olkiluoto site in south-western Finland. The methodology to prepare the safety case documentation will consider the latest updates in the regulations; lessons learned from Posiva s previous safety case, TURVA-2012, submitted in the context of the construction licence application (CLA) in 2012; the feedback received from the Radiation and Nuclear Safety Authority (STUK) on the CLA, including several specific requirements for the next safety case; and new challenges related to the implementation of repository construction and operation. This calls for a higher level of maturity in both the safety case itself and in the design on which the safety case is based. Since the safety case work will inevitably take several years, it is necessary to introduce requirements, design and data freezes at the beginning of the safety case production process. The design freeze is based on the information and requirements available at the start of the safety case work, but updates can be expected as the design matures and is optimized for industrialization and operation. A change management process is set up to facilitate the assessment of the impacts of the proposed changes on the safety case results. The input data used in the safety assessment and their possible updates will be managed by using of a central database. The uncertainties in the initial state of the components of the disposal facility will be tackled by implementing an analysis of potential deviations in these components at the time of installation. Deviations are then screened and implemented in scenario formulation. Defining a range of initial state parameter values and deviations for the installed components introduces some flexibility in design and increases the robustness of the safety case. Key Words: Spent nuclear fuel repository, long-term safety, safety case, Olkiluoto 1. Introduction Posiva Oy is responsible for the disposal of spent nuclear fuel from the Finnish nuclear power plants of Loviisa and Olkiluoto. In November 2015, the Finnish Government granted a construction licence for Posiva s disposal facility at Olkiluoto, in south-western Finland. The construction licence application was supported by a safety case, TURVA-2012 [1], which was evaluated by the Radiation and Nuclear Safety Authority (STUK). STUK concluded that Posiva had developed a safety concept that is in line with the regulatory requirements [2] and that the post-closure safety of the disposal facility has been analyzed in a sufficient manner for the purposes of the construction licence stage [3]. At the moment, Posiva is in the process of preparing a safety case to support the operating licence application (OLA) for the disposal facility. Before the application can be submitted, Posiva will have to fulfil 34 requirements formulated by STUK for the new safety case and the related research and modelling work [3]. The new safety case will also need to consider any updates in the regulations, as well as new challenges related to the implementation of repository construction and operation. 2. Overall Safety Case Methodology A safety case is the synthesis of evidence, analyses and arguments that quantify and substantiate the claim that the repository will be safe after closure and beyond the time when 11

12 active control of the facility can no longer be assumed [4]. A safety case includes a quantitative and a qualitative assessment of the long-term performance of the disposal system. The quantitative assessment (a.k.a. safety assessment) is defined as the process of systematically analyzing the ability of the disposal facility to provide the safety functions and to meet the requirements and of evaluating the potential radiological hazards and compliance with the safety requirements. The qualitative assessment broadens the scope of the safety assessment to include the compilation of a wide range of evidence and arguments that complement and support the reliability of the results of the quantitative analyses [5]. The general safety case structure builds upon the one used in TURVA-2012 [1], i.e. the safety case will consist of a portfolio of main reports and a number of supporting reports. 2.1.Handling uncertainties in the initial state Design development work is moving towards implementation stage and, accordingly, Posiva has planned the disposal operation at a very detailed level, both in order to plan and optimize the disposal operation, but also for production and large-scale implementation tests. The experience obtained to date is used in the safety case to better constrain the uncertainties related to the initial state of the repository system. Initial state refers to the description of the state of various repository components after emplacement has been completed, i.e. information which acts as a starting point for the performance and safety assessments. The uncertainties are handled through a systematic screening of the possible deviations through a modified failure mode and effect analysis (FMEA [6]), and further handling in the scenario formulation work incorporating the deviations into the safety case. The FMEA for the initial state has been modified to screen events that can lead to failure modes that are likely to be undetected and thus remain in the repository at the time of the initial state. The aim is to improve the description of the initial state of the repository system from the traditional design freeze description [7] towards a description of the repository in as-built state. 2.2.Handling uncertainties during the long-term evolution Uncertainties during the long-term evolution of the disposal system are handled through a systematic analysis of how the different FEPs might act on the components of the disposal system during its evolution, followed by the formulation of scenarios and analysis of cases giving rise to potential failures of containment and radionuclide releases and their corresponding radiological impacts. 3. Methodology to Handle Changes 3.1.Requirements, Design and Data freezes Since the safety case work will inevitably take several years, it is necessary to introduce requirements, design and data freezes at the beginning of the safety case production process. The requirements freeze allows the design to be fixed for specific purposes, such as the safety case or large-scale tests. The design freeze is based on the information available at the time of requirements freeze. The data freeze refers to data other than actual design data and includes, for example, geological site data, surface environment data or time-dependent data needed in the modelling chain, where the output of certain models will serve as input to other models. The data freeze does not need to happen at the same time as the design and requirements freeze, only at the time it is needed as input in the modelling chain. Once input to the safety 12

13 Session 3c ILW case has been approved and frozen, its documentation and change management process (see below) is of utmost importance to ensure traceability and reliability of the results in the safety case. The input data will be stored in an electronic central database in a traceable manner, so that both the approved data, approval process and future potential updates are clearly recorded. 3.2.Design Development During the Safety Case Process Requirements, design and data freezes were already used in TURVA-2012 (see, e.g., [7, 8]). One of the lessons learnt was that it is not possible to freeze the design completely before the start of the safety case work, because important developments can happen during the safety case process, which lasts several years while the design develops and operational experience is being obtained. This is expected to be emphasized in the operating licence application process as the design reaches full maturity and is optimized for industrialization and operation. The long-term performance of the design solution as well as further operational aspects, particularly related to the installation of engineered barriers in repository-like conditions will also be studied in large-scale demonstration tests. In their Review Report [2], STUK has concluded that although there are no direct requirements for demonstrations in any of the regulations, the Guide YVL B.1 [9] states that the solutions and methods chosen during the course of the design shall be based on proven technology and operating experience. In addition, the design shall strive for simplicity and, if new solutions are proposed, they shall be validated through tests and experiments [2]. Posiva s plans for large-scale demonstrations are described in the waste management programme YJH-2015 [10]. 3.3.Change Management Process As changes to the design and to other input data may be expected to take place during the safety case work, a change management process needs to be set up to manage the traceability and reliability of the safety case and to facilitate the assessment of impact of changes in design on the safety case results. For this purpose, the whole modelling chain used in the safety case is documented and linked to the approved input data. Configuration management defines the general process to be followed in order to implement a change in the design or requirements for the disposal facility. The heart of the configuration management process consists of classifying each proposed change according to its impact on operations and safety (including long-term safety). Posiva is currently developing the methodology to assess the long-term safety impact of proposed changes within the configuration management process. The criteria to be followed will address the impact of a given design, requirement or process change on the initial state, on the fulfillment of the long-term safety functions, or the overall uncertainty management. Change management is based on expert judgment and relies on a close co-operation between long-term safety and design from the beginning of the safety case work. The main interfaces between these two groups of experts are the long-term safety requirements and their verification as well as the initial state. 4. Conclusions Requirements, design and stepwise data freezes need to be performed in a safety case that is developed in parallel with design optimization and operational readiness activities. A safety case supporting the operating licence application needs a higher level of design maturity than that supporting the construction licence application. In Posiva s case, the design is currently 13

14 being optimized for industrialization and operation and large-scale demonstrations are also taking place, the handling of changes arising from these activities is a major challenge in the safety case process. As some uncertainties in the initial state of the repository components can be assumed to remain, an analysis of potential deviations in these components at the time of installation is proposed to be implemented. The uncertainties in the initial state can then be taken into account in the formulation of scenarios. A change management process needs to be set up to incorporate changes in a controlled way, so that their long-term safety impacts are properly assessed. The proposed changes need to be considered holistically, including the impact not only on long-term safety but also on the safety case production process. The proposed changes will only be accepted if they do not compromise long-term safety and if the safety case analyses can be updated using the new input. Considering the long operational phase (over 100 years) of the disposal facility, further optimization activities are expected to occur as the operational experience and knowledge bases develops; a change management process is thus needed also after the operations have started. 5. References [1] POSIVA OY, Safety Case for the Disposal of Spent Nuclear Fuel at Olkiluoto Synthesis 2012, POSIVA , Eurajoki (2012). [2] RADIATION AND NUCLEAR SAFETY AUTHORITY, STUK s Review on the Construction License Stage Post Closure Safety Case of the Spent Nuclear Fuel Disposal in Olkiluoto, STUK-B 197, Helsinki (2015). [3] RADIATION AND NUCLEAR SAFETY AUTHORITY, Safety Case for the Disposal of Spent Nuclear Fuel in Olkiluoto: Decision, Presentation Memorandum, 1/H42252/2015, Helsinki (2015). [4] NUCLEAR ENERGY AGENCY, Post-Closure Safety Case for Geological Repositories: Nature and Purpose, Report 3679, Paris (2004). [5] POSIVA OY, Safety case plan 2008, POSIVA , Eurajoki (2008). [6] STAMATIS, D.H., Failure Mode and Effect Analysis: FMEA from Theory to Execution. ASQ Quality Press (2003). [7] POSIVA OY, Safety Case for the Disposal of Spent Nuclear Fuel at Olkiluoto Description of the Disposal System 2012, POSIVA , Eurajoki (2012). [8] POSIVA OY, Safety Case for the Disposal of Spent Nuclear Fuel at Olkiluoto Design Basis 2012, POSIVA , Eurajoki (2012). [9] RADIATION AND NUCLEAR SAFETY AUTHORITY, Safety Design of a Nuclear Power Plant, Guide YVL B.1, Helsinki (2014). [10] POSIVA OY, YJH-2015 Nuclear waste management at Olkiluoto and Loviisa power plants: Review of current status and future plans for (in Finnish), YJH-2015, Eurajoki (2015). 14

15 Session 3c ILW 03d 03 / ID 145. Disposal of High Level Waste ANDRA S SAFETY OPTIONS OF FRENCH UNDERGROUND FACILITY CIGÉO A MILESTONE TOWARDS THE LICENSING APPLICATION S. Voinis, M. Rabardy, L. Griffault Andra, French National Radioactive Waste Management Agency, Parc de la Croix Blanche, Châtenay-Malabry, France contact of main author: sylvie.voinis@andra.fr Abstract. Following the publishing of the Dossier 2005 Argile, the 28th June 2006 Act entitled Programme National de Gestion des Matières et Déchets Radioactifs (National program for radioactive waste and nuclear material management) [5] has set the deep geological repository in clay host rock as the selected solution for IL- LL and HL radioactive waste disposal in France. According to this 2006 Act, reversible waste disposal in a deep geological formation and corresponding studies and investigations shall be conducted with a view to selecting a suitable site and to designing a repository. Since 2011, the project has entered an industrial design development phase and has become the Industrial Center for Geological Disposal Cigéo. In view of the licensing application, two main milestones for safety are identified for Cigéo: a Safety Options Dossier DOS early 2016 and the safety case to support Future License Application of Cigéo RPs in According to the 2007 French Act, the Safety Options is an opportunity for the operator to send in advance a first safety case to the French Safety Authority in order to stabilize the safety strategy, the safety requirements, the safety methods, key safety and design options, the list of safety scenario that are selected and a preliminary impact of a few margin scenarii. The Safety options don t present the overall safety demonstration that needs to be presented in the safety case supporting the licensing application. The Cigeo geological disposal facility project is designed to cater for all the HLW and ILWLL that has been produced and will be produced by existing nuclear facilities. Andra has conducted in the frame of the safety options a parallel and coordinated operation and post-closure safety analysis. Those safety options take into account the particularity of Cigéo: HLW and various types of ILWLL waste; the step by step development of Cigeo and the balancing approach between safety, technology and economics. Considering the various families and nature of the ILWLL waste, the Safety Options consist in establishing dimensioning characteristics for design and operational safety as well as envelop inventory to manage the knowledge acquired at this stage. In addition, the classification of scenario and the safety approach are adapted to the operational and post-closure context. The safety options identify the links between the two phases. Key Words: nuclear safety, disposal, waste management, safety options 1. Introduction The purpose of the Cigeo geological disposal facility for HLW and ILW-LL is to allow the safe disposal of IL-HL LL radioactive waste in order to eliminate or reduce the burden to be borne by future generations, in accordance with Article L542-1 of the Environment Code. Since 1991, successive safety milestones were implemented, based on the acquisition of scientific and technical knowledge and the development of safety methods appropriate to deep geological disposal. Since 2011, Cigéo has entered an industrial design development phase. In view of the licensing application, as a key milestones for safety, the Andra Safety Options Dossier DOS early 2016 is submitted to a national review and an international 15

16 review 1. According to the 2007 French Act, the Safety Options is an opportunity for Andra to send in advance a first safety case and aims to stabilize the safety strategy, the waste inventory, safety requirements, the safety methods, key safety and design options, the list of safety scenarios and a preliminary impact of a few margin scenarios. The safety options apply to the disposal of high-level waste (HLW) and intermediate-level long-lived waste (ILW-LL). The Safety options don t present the overall safety demonstration that will be presented in the safety case to support Future License Application of Cigéo in 2018 according to the recent French Act of July FIG. 1. a step-by-step iterative process as regards safety since Safety options and the incremental development of Cigéo The safety options consider the duration of operation for over a hundred years with successive phases (construction/operation); it has to be flexible enough to adapt to possible changes in France's energy policy. There are three main phases in the life of Cigeo: (i) an initial design and initial construction phase, (ii) an operation phase (including an industrial pilot phase) and (iii) a post-closure phase. Cigeo will be closed in stages and the post-closure phase will begin when the final closure of Cigeo has been authorised by a law. FIG 2. Diagram showing the main phases in the life of Cigeo Following final closure, the underground facility after its final closure, will be the facility as built. At the stage of the safety options, the underground layout is only an illustration of what Cigéo might be, based on the technical options chosen at this stage. According to the 1 International review by expert from regulatory and IAEA on behalf of the French Safety Authority 16

17 Session 3c ILW incremental development of Cigéo, if a new technological solution is suggested, it will be checked that the operational and post-closure safety functions are still fulfilled (safety indicator assessment) and the radiological impact remains as low as reasonably possible given economic and social factors. 3. The disposal system (natural and engineered components) relies on both operational and post-closure safety principles and safety functions Protecting people and the environment is primarily based on the performance of safety functions during operation comparable to those performed at all nuclear facilities, and on safety standards (national and international), safety requirements and safety options adapted to the specific underground environment of the Cigéo facility. Andra has implemented, from the design stage (since the 90 s), a safety approach and process (including siting), which relies on the specific characteristics of a repository as such: (i) the choice of the Callovo-Oxfordian formation in which the underground facility of Cigeo is located that meets the site technical criteria of the 2008 ASN nuclear safety guide, (ii) an underground facility located at a depth of around 500 m, of reduced geometry and long connecting drifts, requiring specific operating, intervention and evacuation conditions; (iii) a disposal facility being developed in successive phases, implying a need to consider the risks related to performing underground construction work and nuclear operations in parallel; (iv) a coordinated approach encompassing operating safety and post-closure safety. The approach integrates the successive iterations of Cigéo milestones including design and scientific knowledge evolutions with the objective of ensuring post-closure safety throughout the entire development cycle of the Cigeo project. An appropriate level of monitoring and control will be also applied to Cigeo from its construction and during its operation, to ensure the protection and preservation of the passive post-closure safety features, as necessary, so that they can fulfil their safety functions once the repository is closed. During operational phase, five safety functions are applicable to Cigeo throughout the operating phase and must be maintained in all incident or accident situations of internal or external origin or, at least, restored within time limits consistent with the objectives of protecting people and the environment defined for the Cigeo project. They are: (i) contain radioactive substances to protect against the risk of their dispersion; (ii) protect people from exposure to ionising radiation; (iii) manage safety with regard to the criticality risk; (iv) remove the heat produced by waste and (v) remove gases formed by radiolysis in order to manage explosion risks. For the post-closure phase, the Cigéo aims to isolate the waste from humans and the biosphere and to confine it within a deep geological formation to prevent dissemination of the radionuclides contained in this waste (see table 1). The post-closure disposal system relies particularly on the Callovo-Oxfordian that plays the main role, and the closure structures of the surface-to-bottom connections (sealed shafts and ramps). The global approach to postclosure safety assessment is based on practical expression of the safety functions and associated requirements, analysis of component performance and analysis of the uncertainties related to the scientific and technological knowledge underpinning the design. To fulfil the post-closure safety functions, design principles for Cigeo and for the choice of site (see examples in table 1) are established. 17

18 TABLE I: Example of Safety Principles for Cigéo Post-closure safety functions General principles in terms of choice of site and design Isolating waste from surface phenomena and human actions Location of Cigeo at a depth and in an area of low, uniform geodynamic Preventing the circulation of water Restricting the release of radionuclides and toxic elements and immobilising them in the repository Delaying and reducing radionuclide migration Low water flows in the Callovo-Oxfordian due to its low permeability and the low hydraulic head gradient; Consolidation and sealing of the surface-to-bottom connections Cells (particularly the materials used for them) designed to protect the waste and packages from a physicochemical point of view Thickness of Callovo-Oxfordian (at least 130 m), high retention capacity Optimised geometries of the cells and drifts in the underground installation, particularly in terms of length. Whether the disposal system is functioning correctly and more specifically whether the safety functions are being fulfilled (operation normal functioning and post-closure normal evolution,) the design options relies also on the results of the risks analysis during operational phase adapted to Cigéo context (mainly transfer of waste package, co-activity of works and operation..) and the subsequent scenario (e.g. dimensioning waste characteristics and scenarios). It also relies on the scientific and technological uncertainties analysis after closure and the resulting normal evolution, altered evolution and what-if scenarios assessment. In the case of Cigéo, the safety options present a series of scenarios considering the dysfunction of sealing, the dysfunction of vitrified waste canister, as well the occurrence of inadvertent human intrusions (mostly borehole for Cigéo). Quantitative evaluations aimed at considering envelop situations of those scenarios. FIG 3. Diagram showing the coordinated approach to operating safety and post-closure safety 18

19 Session 3c ILW REFERENCES [1] Act of 30 December 1991 on radioactive waste management research. (1992). Official Journal of the French Republic Acts and Decrees No. 1, 10 p. [2] Act of 28 June 2006 on the sustainable management of radioactive material and waste. (2006). Official Journal of the French Republic. Acts and Decrees No. 93, 9,721 p. [3] Délibération du conseil d'administration de l'agence nationale pour la gestion des déchets radioactifs du 5 mai 2014 relative aux suites à donner au débat public sur le projet CIGEO. Ministère de l'écologie, du développement durable et de l'énergie (2014). Journal Officiel. Lois et décrets, n 108, pp Safety Options Report Post-Closure Part (DOS-AF). Andra. (2015). CGTEDNTEAMOASR [4] Safety Options Report Operation Part (DOS-Expl). Andra. (2015). CGTEDNTEAMOASR [5] Act No of 13 June 2006, as amended, on transparency and security in the nuclear field. Consolidated version dated 12 July (2006). [6] Order of 7 February 2012 laying down the general rules on basic nuclear installations Consolidated version dated 05 July (2012). [7] NEA IGSC Scenario Development Workshop, 1-3 June 2015, Issy-les-Moulineaux, France, to be published, OCDE. [8] Radioactive Waste Disposal Facilities Safety Reference Levels v2.2. (Wgwd), W.G.O.W.a.D. Western European Nuclear Regulators Association (WENRA). (2014). 81 p. [9] Fundamental safety principles. Safety fundamentals. IAEA. (2006). IAEA safety standards series n SF p. [10] Disposal of Radioactive Waste. Specific Safety Requirements. IAEA. (2011). IAEA Safety Standards Series n SSR p. [11] The management system for facilities and activities. Safety Requirements. IAEA. (2006). IAEA Safety Standards Series n GS-R p. [12] Monitoring and Surveillance of Radioactive Waste Disposal Facilities. Specific Safety Guide. IAEA. (2014). IAEA Safety Standards Series n SSG p. [13] The Safety Case and Safety Assessment for the Disposal of Radioactive Waste. Specific safety guide. IAEA. (2012). IAEA Safety Standards Series n SSG p. [14] Geological Disposal Facilities for Radioactive Waste. Specific Safety Guide. IAEA. (2011). IAEA Safety Standards Series n SSG p. [15] The management system for the disposal of radioactive waste. Safety guide. IAEA. (2008). IAEA safety standards series n GS-G p. 19

20 03d 04 / ID 172. Disposal of High Level Waste BRIDGING NUCLEAR SAFETY, SECURITY AND SAFEGUARDS AT GEOLOGICAL DISPOSAL OF HIGH LEVEL RADIOACTIVE WASTE AND SPENT NUCLEAR FUEL I. Niemeyer, G. Deissmann, D. Bosbach Forschungszentrum Jülich GmbH, IEK-6: Nuclear Waste Management and Reactor Safety contact of main author: i.niemeyer@fz-juelich.de Abstract. In order to consider geological disposal of high-level radioactive waste and spent nuclear fuel in all its complexity, related nuclear safety, security and safeguards issues have to be taken into account. By identifying both synergies in overlapping methods or techniques and differences in the requirements with respect to safety, security and safeguards, advantage of inherent synergies and conflicting requirements can be taken at the same time. While there is a general understanding of the potential benefits of the 3S concept, neither the interfaces and synergies between safety, security and safeguards nor their practical implementation are yet fully understood. This paper discusses the role and importance of safety, security and safeguards regarding the geological disposal of high-level radioactive waste and spent fuel. Key Words: Safety; security; safeguards; 3S 1. Introduction The use of the terms nuclear safety, nuclear security and nuclear safeguards is often not sharply delimited from each other, though definitions for each of these issues exist. According to IAEA definitions, nuclear safety refers to [t]he achievement of proper operating conditions, prevention of accidents or mitigation of accident consequences, resulting in protection of workers, the public and the environment from undue radiation hazards [1], and therefore stands for the safe operation of nuclear installations. Nuclear security implies [t]he prevention and detection of, and response to, theft, sabotage, unauthorized access, illegal transfer or other malicious acts involving nuclear material, other radioactive substances or their associated facilities [1] and is aimed at the physical protection of nuclear installations. Nuclear safeguards are designed to ensure that special fissionable and other materials, services, equipment, facilities, and information made available by the Agency or at its request or under its supervision or control are not used in such a way as to further any military purpose [2] or, in short, to ensure the peaceful use of nuclear material. The interaction or intersections of the three components depend on the context, and the significance of each of the components may vary for different types of nuclear installations. In order to consider geological disposal of high-level radioactive waste and spent nuclear fuel in its full complexity, all related nuclear safety, security and safeguards issues must be taken into account. While safety can benefit from some provisions regarding safeguards and physical protection (security), it may also be contravened by others. Some techniques for monitoring geological repositories, such as environmental sampling, could provide relevant data for safety, security and safeguards. Other techniques, such as geophysical measurements for safeguards verification, are to be implemented in a way that does not infringe long-term safety requirements. Therefore, identifying both synergies in overlapping methods or 20

21 Session 3c ILW techniques or with respect to their future development as well as differences in the requirements with respect to safety, security and safeguards may help to take advantage of inherent synergies and conflicting requirements at the same time. The need of integrating the three S s, also referred to as the 3S concept, to the extent possible throughout all the stages of the nuclear installations life cycle, was recognized by the IAEA in 2008 [3,4] and at the same time, the G8 countries declared to support the 3S concept [5,6]. Since then, a number of papers discussed the benefits of considering a 3S approach [e.g., 7-9] in designing and operating nuclear facilities, but only a few addressed the issue of applying 3S to geological disposal of high-level radioactive waste and spent nuclear fuel [10-12]. While there is a general understanding of the potential benefits of the 3S concept, neither the interfaces and synergies between safety, security and safeguards nor their practical implementation are fully understood to date. This also applies to the geological disposal of high-level radioactive waste and spent nuclear fuel. Numerous legislations, regulations and other documents have emphasized that safety is the primary requisite in all life cycle stages of geological repositories. But what is the significance of security and safeguards with respect to geological disposal? 2. Role and importance of safety, security and safeguards regarding the geological disposal of high-level radioactive waste and spent nuclear fuel 2.1.Legal and organizational framework Nuclear safety, security and safeguards legislations are laid down in a series of national and international agreements, conventions and regulations [13]. With reference to the 3S concept, the IAEA noted the need for nuclear legislation that reflects the interrelations between safety, security and safeguards in a comprehensive and synergetic manner [14]. Accordingly, any new or revised nuclear legislation on geological disposal of high-level radioactive waste and spent nuclear fuel should also take 3S conflicts and interfaces into account. Safety and security are mainly based on an appropriate national legal and organizational framework, including national regulatory oversight of safety and national law enforcement in case of security threats. Safeguards, however, represents an international legal commitment, determined by safeguards agreements and additional protocols between States and the IAEA [15]. States under safeguards verification by the IAEA usually have a national or regional Safeguards Regulatory Authority (SRA) in place that acts as interface between the State and the IAEA. Some States, such as Finland and Japan, have established national regulatory bodies that cover safety, security and safeguards issues of their nuclear installations and programmes, including geological disposal, in a single organization [10]. 2.2.Material concerned Safe geological disposal requires a stable geological formation to provide for the long term containment of radionuclides and their isolation from the biosphere. Safety therefore addresses all types of radionuclides, in particular the long-lived ones (with half-life periods in the order of up to 10 7 years), i.e. actinides and long-lived fission and activation products. Security considers nuclear material and other radioactive material [1], and safeguards are principally applied to all source (uranium, thorium) or special fissionable material containing uranium or plutonium [2]. The lowest common denominator of a 3S control of nuclear 21

22 material in high-level waste and spent nuclear fuel would therefore include uranium, plutonium and thorium. 2.3.Timelines The safety case and safety assessment for geological disposal facilities consider the three life cycle stages, i.e. the pre-operational period, the operational period and the post-closure period, spanning over periods in the order of thousands of years and potentially longer (i.e. up to hundreds of thousands of years) [16]. Security measures do address the three life cycle stages as well, with a focus on the pre-operational and operational periods, although a generally care and maintenance free post-closure phase is stipulated in the regulations in various countries. The timeline for safeguards activities is bound by the duration of the safeguards agreements and, in the end, will be applied as long as the Nuclear Nonproliferation Treaty (NPT) remains in force. A 3S assessment should thus be based on the longest timeline of the single S -components, while the role and importance of each of the three S s would vary or decrease over time. If a 3S case was to be prepared instead of the safety (1S) case, the long-term post-closure period would mainly be assessed from the safety perspective. 2.4.Control measures Safety, security and safeguards activities include similar or complementary measures for documenting, measuring and monitoring the inventory of radionuclides, in particular with regard to uranium, plutonium and thorium. In order to avoid redundancy or duplication of work and equipment, a material control and accountancy system should include practices and procedures, as well as techniques for measurement, sealing and surveillance that fulfil the requirements as to safety, security and safeguards to the extent possible. 2.5.Facility design The IAEA generally considers safety, security and safeguards as essential elements in all life cycle stages of nuclear facilities. In this context, the IAEA has issued a guidance document [17] aimed at informing stakeholders how to design facilities for nuclear waste management by early consideration of safeguards in the planning stage so that provisions can be better integrated with other design requirements as to safety and security. This approach, also referred to as safeguards by design (SBD) should be more closely interlocked with the 3S concept. Safety, security, safeguards by design (3SBD), as generally proposed by [18,19], can help to reduce efforts and costs related to nuclear waste management and disposal. 3. Findings Safety, security and safeguards aspects regarding the geological disposal of high-level radioactive waste and spent fuel should be addressed and managed in a coordinated, complementary approach. Further R&D will be needed to identify methods and technologies (a 3S toolbox ) that would be best suited for the holistic consideration of safety, security and safeguards provisions. By early consideration of conflicting requirements as to safety, security and safeguards, their impacts on all three life cycle stages of geological disposal can be minimized. The 3SBD toolbox should include methods and technologies for material accountancy, nuclear measurements, containment and surveillance, environmental 22

23 Session 3c ILW monitoring, continuity of knowledge, as well as design implications to the benefit of all safety, security and safeguards at geological disposal. REFERENCES [1] INTERNATIONAL ATOMIC ENERGY AGENCY, IAEA Safety Glossary. Terminology used in Nuclear Safety and Radiation Protection, Vienna (2007). [2] INTERNATIONAL ATOMIC ENERGY AGENCY, IAEA Safeguards Glossary, Vienna (2001). [3] INTERNATIONAL ATOMIC ENERGY AGENCY, 20/20 Vision for the Future, Background Report by the Director General for the Commission of Eminent Persons, Vienna (2008). [4] INTERNATIONAL ATOMIC ENERGY AGENCY, Reinforcing the Global Nuclear Order for Peace and Prosperity Role of the IAEA to 2020 and Beyond, Vienna (2008) [5] G8 HOKKAIDO TOYAKO SUMMIT LEADERS DECLARATION, World Economy, Paragraphs 28 and 65, Hokkaido Toyako, Japan (2008). [6] TSUTOMU, A., NAITO, K. The New Nexus, 3S: Safeguards, Safety, Security, and 3S- Based Infrastructure Development for the Peaceful Uses of Nuclear Energy, Journal of Nuclear Materials Management (JNMM) 34(4) (2012), [7] KIM, H., et al., 3S (Safety, Security, and Safeguards)-by-Design for Engineering-Scale Pyroprocessing Facility, Proc. ESARDA Annual Meeting, 35th Annual Meeting, Bruges (2013). [8] LEE, N.Y., et al., 3S Culture, Its Meaning and Future Direction, Proc. INMM 55th Annual Meeting, Atlanta, GA (2014). [9] SANDERS, K.E., et al., Interfaces among Safety, Security, and Safeguards (3S) - Conflicts and Synergies, Proc. INMM 56th Annual Meeting, Indian Wells, CA (2015). [10] VAJORANTA, T., Finland s Integrated Approach to Safety, Security, and Safeguards, IAEA Technical Meeting on Safety, Security and Safeguards, Vienna (2012). [11] MARTIKKA, E., et al., Safeguards for a Disposal Facility for Spent Nuclear Fuel a Challenge for 3S, Proc. INMM 55th Annual Meeting, Palm Desert, CA (2013). [12] HADDAL, R., et al., Geological Repository Safeguards: Options for the Future, Proc. IAEA Symposium on International Safeguards: Linking Strategy, Implementation and People, Vienna (2014). [13] INTERNATIONAL ATOMIC ENERGY AGENCY, Handbook on Nuclear Law, STI/PUB/1160, Vienna (2003) [14] INTERNATIONAL ATOMIC ENERGY AGENCY, Handbook on Nuclear Law: Implementing Legislation, STI/PUB/1456, Vienna (2010). [15] CHERF, A., Legal Framework for Safety, Security and Safeguards, IAEA Technical Meeting on Safety, Security and Safeguards, Vienna (2012). [16] INTERNATIONAL ATOMIC ENERGY AGENCY, The Safety Case and Safety Assessment for the Disposal of Radioactive Waste, IAEA Safety Standards Series No. SSG-23, Vienna (2012). 23

24 [17] INTERNATIONAL ATOMIC ENERGY AGENCY, International Safeguards in the Design of Facilities for Long Term Spent Fuel Management, IAEA Nuclear Energy Series No. NF-T-3.1, Vienna (in print). [18] STEIN, M., MORICHI, M., Safety, Security, and Safeguards by Design: An Industrial Approach, ANS Nuclear Technology 179(1) (2012) [19] NUCLEAR DECOMMISSIONING AUTHORITY, Geological Disposal Safety, Environmental, Security and Safeguards Principles for the Design Process, NDA Technical Note no

25 Session 3c ILW 03d 05 / ID 141. Disposal of High Level Waste DEVELOPMENT OF THE NUMO PRE-SELECTION, SITE-SPECIFIC SAFETY CASE T. Fujiyama, S. Suzuki, A. Deguchi, H. Umeki Nuclear Waste Management Organization of Japan (NUMO), Tokyo, Japan contact of main author: tfujiyama@numo.or.jp Abstract. NUMO has developed a safety case for co-disposal of HLW and TRU waste to reflect current boundary conditions in Japan. In particular, this involves addressing public concerns in the wake of the Fukushima Dai-ichi nuclear power plant accident and a move by the Government to more strongly support moving forward with siting a geological repository, involving suggesting locations that are considered to be scientifically more suitable. This paper will provide a brief overview of this Safety Case, with a focus on advances from the old H12 Report, which is considered the first generic safety case in Japan. The NUMO pre-selection, site-specific safety case has been developed to provide a basic structure for subsequent safety cases that would be applied to any selected site, emphasising practical approaches and methodology, which will be applicable for the conditions/constraints during an actual siting process. Key Words: Geological disposal, Safety case, Vitrified waste, TRU wastes. 1. Introduction The H12 Report [1] published in 1999 by the Japan Nuclear Cycle Development Institute (now the Japan Atomic Energy Agency, JAEA ) demonstrated the feasibility of safe and technically reliable geological disposal of high level waste (HLW), based on a generic study. On the basis of the H12 Report, the Final Disposal Act came into force and NUMO was established as the implementing body in Intermediate-level waste generated from reprocessing of spent nuclear fuel and mixed-oxide fuel fabrication (termed TRU waste in Japan) was also included in the inventory for disposal in NUMO has been developing key technologies required for the safe implementation of the geological disposal project since its establishment and initiated the siting process by open solicitation of volunteer municipalities in So far, however, no volunteer municipality has appeared and no candidate host rock type can be specified. The Fukushima-Daiichi NPP accident in 2011 increased nationwide concerns about the feasibility and reliability of safe geological disposal in Japan. After re-thinking the implementation process, the Basic Policy on the Final Disposal of Specified Radioactive Waste was amended in 2015, so that the Government now leads the search for volunteer sites. This procedure involves nominating more suitable areas from a geo-scientific point of view to initiate discussions and cooperation with local municipalities, finally leading to acceptance of a site investigation, which will be carried out by NUMO. Taking such changes in boundary conditions into account, it is important at this time for NUMO to integrate required technologies, including the latest R&D output, in order to confirm the feasibility and safety of geological disposal in Japan. Thus NUMO has developed the NUMO pre-selection, site-specific safety case which, with the site descriptive models (SDMs) recently developed, provides a more advanced site-specific basis than the generic safety case in the H12 Report. This has been developed in a manner that will allow it to provide the basic structure of future safety cases that would be applicable to any site that might arise from the site selection process. 25

26 3 km 100m Session 3d HLW 2. Basic Strategy of the NUMO Safety Case Despite the fact that there has not been a site or specific host rock identified, the NUMO Safety Case has developed detailed geological and hydrogeological models for potential host rock environments. Repository design and safety assessment have been thus performed for these geological models, thereby providing underpinning evidence to demonstrate the technical feasibility and the safety for the various types of Japanese geological environments. More background is provided in the companion paper by Suzuki et al. [2] 3. Site Characterisation and Synthesis into SDMs NUMO needs to prepare reliable investigation and evaluation methodologies and an approach to synthesise their output in order to form the basis of selecting suitable repository sites. In suitable setting, the key safety functions (isolation and containment) of the host geological environment will persist for a long period of time. Advanced methodologies for precluding the potential impacts of natural disruptive events and processes are shown. Key concepts, technical knowledge bases, and basic methodology for geosynthesis of relevant information into representative spatial and temporal models of site evolution are also documented. The illustrative site descriptive models (SDMs) are developed for subsequent repository design and safety assessment. Generic repository design and safety assessment were performed in the H12 Report for two illustrative geological settings, namely crystalline rock and sedimentary rock. However, geoscientific knowledge has expanded significantly since then due in particular to multidisciplinary research programmes; for example, JAEA s underground research laboratory projects. It is thus very important to update the previous repository design and safety assessment on the basis of the current best understanding of Japanese geological environments in the NUMO Safety Case. Following the categorization of all potential host rock environments, rock types are grouped by identifying key characteristics/properties relevant to geological disposal. As a result, three types of potential host rock environments, Igneous rocks, Neogene sedimentary rocks, and Pre-Neogene sedimentary rocks are examined in the NUMO Safety Case. Highly fractured (weathered) domain Sedimentary overburden Regional scale 50 km 50 km 3 km Granite GW flow 1km Active fault Illustrative geological setting 100~200m 100~200m Active fault Unpreferable area Faults (Length > 1km) Short travel time Approx. 800m 100m Reserved area Reserved area Reserved area Repository scale Panel scale EBS + Rock Near-field scale FIG. 1. An example of the nested SDMs for Igneous rocks, including underground panel layout (bottom, left) and EBS of HLW (bottom, right) Realistic geological and hydrogeological models are developed in a stepwise manner for the three types of potential host rock environments: at scales of several kilometres (repository 26

27 Session 3c ILW scale), for defining the location and layout of a repository and assessing groundwater flow through the potential host rock; then at several hundred metres (panel scale) and a hundred metres (near-field scale), for more precisely describing hydraulic properties. FIG. 1 shows an example of the geological and hydrogeological modelling for Igneous rocks at nested scales. For geological modelling, key geological structures that control groundwater flow and have a major influence on solute transport, such as faults, fractures and sedimentary structures, are represented by a combination of deterministic and stochastic modelling approaches. 4. Repository Design Design methodologies should be developed so as to maintain flexibility for the range of potential geological conditions encountered in Japan. In the NUMO Safety Case, alternative repository concepts are presented, which are applicable for a wide range of potential geological conditions. The designed repositories should be technically feasible to construct and fulfil the safety functions required to isolate and contain radionuclides. The design requirements and specifications of the engineered barrier system (EBS), disposal tunnel, panel layout and sealing system (tunnel back-filling and plugs) have been defined. The methodology is demonstrated by carrying out a full repository design study, tailored to the SDMs of three types of potential host rock. The engineering feasibility of construction, operation and closure of the repository is evaluated based on techniques demonstrated in domestic or overseas underground laboratories and related R&D facilities. The diagram in FIG. 1 (bottom, left) illustrates an example of an underground layout tailored to the geological and hydrogeological model for Igneous rocks. The single level emplacement panel is applicable in this case, avoiding faults with lengths greater than 1 km (the minimum length that can be identified by surface-based investigation), and avoiding any less preferable areas where calculated groundwater travel times are relatively short. Optimal operational processes and material flow logistics, ventilation and water drainage systems for the underground facility are also considered while determining the layout. Such site-specific design demonstrations show progress in practicality of design methodologies. 5. Safety assessment During the siting stages, both pre- and post- closure safety will proceed in an iterative manner and the resulting output will support decisions made at the end of each stage from the perspective on safety. The required safety assessment technology for scenario development, modelling, database development, etc. will be maintained at the state-of-the-art. For pre-closure safety, it is needed to demonstrate the feasibility of radiological protection for local residents and workers during repository operation. Learning from the Fukushima- Daiichi NPP accident, the regulatory guidelines for nuclear facilities have been revised, but those for geological disposal have not been discussed in detail so far. In developing methodology for operational safety assessment of geological disposal, relevant guidelines for other nuclear facilities, such as those for vitrified waste storage, are considered in the NUMO Safety Case. An the first stage, radiological safety is highlighted, focusing on activities relevant to HLW handling and transport, based on specific repository designs and defined procedures of operation. For post-closure safety, it is needed to develop an assessment approach and methodologies which can be applied to specific sites and the repository design concepts tailored to them. In the NUMO Safety Case, procedures and methodologies to assess long-term safety are 27

28 demonstrated. A risk-informed approach is introduced, based on international guidelines as well as recent national discussions on safety regulations. Scenarios are developed and classified with consideration of their probability of occurrence and target doses are defined as illustrated TABLE I. Referring to the guidelines of international organizations on assessment timescales, dose calculations are carried out for up to one million years after closure. TABLE I: SCENARIO CLASSIFICATION AND TARGET DOSE Scenario classification Definition Target dose Likely Scenario The scenario is used to assess the performance of the geological disposal system based on best understanding of the probable evolution, as a 10 μsvy -1 reference for the optimisation of protection. Less-likely scenario The scenario is used to assess the safety of the geological disposal system in view of uncertainties in scientific knowledge supporting likely scenarios. 0.3 msvy -1 Very unlikely scenario Human intrusion scenario Possible scenarios with extremely low likelihood msvy -1 The scenario is used to check whether the geological disposal system is robust with assumption of the human intrusion after loss of institutional control. Residents:1-20 msvy -1 Intruder: msv per event A hybrid methodology of scenario development is introduced, which combines a more conventional, bottom-up, FEP-based approach and a top-down method based on safety functions, appropriate to a risk-informed assessment. Safety assessment is being conducted by applying a approach and methodology of realistic radionuclide transport modelling, as needed to allow comparison of sites and also possible repository concepts that could be tailored to them. This advanced modelling includes more realistically representing the geometry of all components of the engineered barriers (essential for distinguishing between different repository design options) and realistically representing the 3-dimensional geometry of the geosphere, with particular emphasis on the solute transport characteristics of all relevant formations (shown in FIG.1, bottom right). The estimated doses of scenarios in different categories are smaller than the assumed criteria in the NUMO Safety Case. This provides a basis for more comprehensive demonstration of post-closure safety at this stage. The outline of assessment pre- and post- closure safety in the NUMO Safety Case is presented in the companion paper. [2] 6. Conclusions and a look to the future The safety case developed by NUMO is inherently limited by the lack of an actual site to focus on, but the SDM-based approach provides critical experience in integrating the activities of site characterisation and engineering design teams, focused by the fundamental requirement to robustly assure safety. This will prove invaluable in the next phase when parallel characterisation of potential sites may occur. REFERENCES [1] Japan Nuclear Cycle Development Institute, H12: Project to establish the scientific and technical basis for HLW disposal in Japan, JNC-TN ~004, (2000). [2] S. Suzuki, et al., Assessment of pre- and post-closure safety in the NUMO safety case for a geological repository, International Conference on the Safety of Radioactive Waste Management, IAEA, Vienna (2016) (in press). 28

29 Session 3c ILW 03d 06 / ID 131. Disposal of High Level Waste DEVELOPMENT OF A GENERIC ENVIRONMENTAL SAFETY CASE FOR THE DISPOSAL OF HIGHER ACTIVITY WASTES IN THE UK L.E.F. Bailey 1, T.W. Hicks 2 1 Radioactive Waste Management, Building 587, Curie Avenue, Harwell, Oxford, Didcot OX11 0RH, UK 2 Galson Sciences Limited, 5, Grosvenor House, Melton Road, Oakham, LE15 6AX, UK contact of main author: Lucy.Bailey@nda.gov.uk Abstract. The UK is committed to implementing geological disposal for the long-term, safe management of higher activity radioactive wastes [1]. Higher activity waste includes low-level waste not suitable for nearsurface disposal, intermediate-level waste and high-level waste. As yet, no site has been selected for a geological disposal facility (GDF) in the UK, but it has been agreed that a site will be sought using a consentbased approach, preceded by a national geological screening process. Radioactive Waste Management Ltd (RWM) is responsible for the delivery of the GDF. RWM also has a role to support the ongoing packaging of radioactive wastes and to provide disposability assessments for waste producers to provide confidence that packaged wastes will be suitable for eventual disposal in the GDF. To underpin its role, it is essential that RWM can demonstrate its confidence in the safety of a GDF. To this end, RWM maintains a generic Disposal System Safety Case that addresses all safety issues concerning the transport of radioactive wastes to a GDF, the construction and operation of a GDF and the long-term, environmental safety of a GDF after it has been sealed and closed. It is a challenge to present a meaningful safety case when the location and hence the design of the GDF are not known. This is particularly pertinent for the long-term, environmental safety case (ESC), which depends significantly upon an understanding of the geological setting of the GDF and its evolution. This paper explains how RWM has developed a generic ESC based on an understanding of the environmental safety functions provided by a multi-barrier disposal system and the features, events and processes (FEPs) that support them [2]. It explains how an understanding of the basic physics and chemistry underpinning generic GDF concepts can be used to develop insight models to build understanding of the long-term performance of a GDF, to support a safety narrative. The paper also explains the role of probabilistic total system models in providing illustrative calculations to support the generic ESC and RWM s approach to addressing the inevitable uncertainties associated with the long timescales that need to be considered within an ESC. Key Words: radioactive, waste, disposal, safety 1. Introduction The environmental safety of geological disposal is achieved by isolating the wastes in a facility constructed deep underground and ensuring that the radionuclides and nonradiological contaminants are contained such that long-term safety is provided by passive means. To support the development of a GDF concept that achieves environmental safety, RWM has defined a set of long-term safety requirements that are consistent with regulatory expectations on environmental safety. In the absence of a GDF site, while geological screening progresses in the UK, the strategy for GDF design to meet the safety requirements is focused on the development of illustrative concepts for radioactive waste disposal in three generic rock types (higher strength rock, lower strength sedimentary rock and evaporite rock). For each host rock type, illustrative GDF designs have been identified for high-heatgenerating wastes (HHGW) and low-heat-generating wastes (LHGW), based on multi-barrier concepts that have been developed in the UK or overseas. 29

30 2. Generic ESC strategy The generic ESC [Error! Bookmark not defined.] focuses on a narrative that presents RWM s understanding of safety in the context of the illustrative disposal concepts and their barrier system components. RWM has defined a general set of environmental safety functions that could be provided by different barrier system components at different times after disposal. These environmental safety functions define how the geological environment, wasteform, container, buffer/backfill, plugs and seals of a GDF combine to isolate and contain the wastes and limit the transport of contaminants to the surface environment in groundwater or gas. They also relate to how the stability of the barrier system is maintained, how disruption of the barrier system through gas-pressurisation is avoided, and how the potential for post-closure criticality is minimised. Each environmental safety function will be influenced by various FEPs after disposal. For RWM s generic ESC, the OECD NEA FEP database [3] was reviewed to identify FEPs of potential relevance to the different GDF concepts and the safety functions provided by the barrier systems. For example, FIG. 1 shows the waste package FEPs listed in the OECD NEA international FEP database that could influence how a wasteform limits the release of contaminants from a waste package. Having defined how different barriers provide environmental safety functions after disposal, the overall environmental safety of a GDF needs to be demonstrated for all relevant scenarios of disposal system evolution. These scenarios are identified through an analysis that considers the timescales over which each barrier s environmental safety functions are expected to be effective and the situations in which barrier performance may be challenged or compromised by probabilistic events and processes. FIG 1 Illustration of the waste package FEPs listed in the OECD NEA international FEP database that could influence how a wasteform limits the release of contaminants from a waste package 30

31 Session 3c ILW By considering barrier behaviour for each illustrative disposal concept, a base scenario has been defined that represents understanding of expected GDF evolution, with reference to the environmental safety functions to be provided by each barrier component. Variant scenarios for GDF evolution have been identified based on consideration of probabilistic FEPs that may or may not occur. By assessing these scenarios, different safety states are analysed in which wastes are contained in the engineered barrier system or in the geological barrier, or are returned in residual amounts to the accessible environment at regulated, acceptable levels Evaluation strategy It is possible to gain understanding of the post-closure performance of a GDF by considering the basic physics of the disposal system through insight modelling. Such modelling can be used to inform the development of appropriate disposal concepts for different generic rock types and is particularly helpful at the generic stage because it does not require large amounts of data. For example, the peak radiological risk arising from the migration of radionuclides via the groundwater pathway in an advection-dominated geological environment can be estimated using a simple one-dimensional model of radionuclide transport in a porous medium. Such a model can be used to illustrate in terms of a set of dimensionless parameters how peak radiological risk is low if there is a long retarded travel time in the geological barrier relative to the rate of radionuclide decay, significant longitudinal dispersion along the transport path through the geological barrier, or slow leaching of radionuclides from the disposal region relative to the rate of radionuclide decay. Insight modelling complements the more detailed probabilistic total system modelling of the behaviour of radionuclides and non-radiological species in a disposal system that takes account of uncertainty in system evolution. These models enable the risks associated with geological disposal, as well as complementary indicators of safety such as activity fluxes across barriers, to be evaluated, thus supporting understanding of the different safety states of a GDF. Natural and archaeological analogue evidence of how barrier materials behave under expected disposal conditions presents other lines of reasoning that support safety evaluations Assessment timescales In considering appropriate assessment timescales it is relevant to consider the hazard presented by the wastes in the GDF and the uncertainties associated with the GDF and its environment. For the period in which a disposal system is expected to be relatively stable and uncertainties in the behaviour of radionuclides and non-radiological contaminants can be quantified more reliably, it is appropriate to undertake probabilistic calculations of GDF performance. However, for periods in excess of a few hundred thousand years after GDF closure, the geological environment could be affected by large-scale natural processes, such as tectonism, subsidence, uplift and erosion, permafrost development and periods of glaciation. Until specific sites have been identified as potential locations to host a GDF in the UK, RWM considers that it is not appropriate to undertake a detailed assessment of the impacts of large-scale natural processes on GDF performance. Thus, the generic ESC includes probabilistic calculations for an assessment period of 300,000 years after GDF closure to provide an indication of the barriers that are likely to be of most importance to GDF performance on the timescales over which some large-scale natural processes could occur. In this period, the activities of the majority of radionuclides are expected to become insignificant as a result of radioactive decay whilst the radionuclides are contained within the disposal system. The behaviour of relatively soluble and mobile radionuclides, such as chlorine and iodine, and gaseous releases of radionuclides are likely to be of most 31

32 significance to environmental safety in this period. A range of illustrative calculations of radionuclide transport and containment have been undertaken to assess base and variant scenarios over the 300,000-year period for the different illustrative disposal concepts. 3. Summary The generic ESC explains how the geological disposal of higher activity wastes can be accomplished in a way that ensures environmental safety in the long term after wastes have been emplaced and the disposal facility has been closed. Underpinning the ESC are: A safety concept that is based on ensuring that the long-term safety requirements for the GDF are met. A demonstration of how environmental safety can be achieved by implementing disposal concepts that are based on systems of multiple engineered and natural barriers that provide multiple safety functions. These barriers are designed to ensure that the wastes are isolated and contained for the long term after disposal by passive means. An understanding of expected barrier performance and how conditions in a disposal system will evolve, based on research findings presented in RWM s knowledge base. An approach to safety assessment based on multiple lines of reasoning, involving both qualitative and quantitative analysis. Insight modelling and total system modelling have been used to develop an understanding of how different components of the engineered and natural barrier system contribute to environmental safety. At the current time, no site is available for a GDF in the UK and therefore the ESC is necessarily generic. The high-level generic safety arguments presented in this ESC provide the understanding that will underpin the future development of a site-specific ESC. In particular, at each stage of the development and design of the GDF, demonstration of the post-closure safety of a disposal concept will be founded on an understanding of the environmental safety functions that will be provided by the specific engineered barriers defined for a particular combination of host rock and wasteform and the natural barriers provided by the geological environment. REFERENCES [1] DEPARTMENT OF ENERGY & CLIMATE CHANGE, Implementing Geological Disposal A Framework for the Long Term Management of Higher Activity Waste, URN 14D/235, July [2] RADIOACTIVE WASTE MANAGEMENT, Generic Environmental Safety Case Main Report, DSSC/203/01, in publication. [3] NEA, Updating the NEA International FEP List: An IGSC Technical Note, Technical Note 2: Proposed Revisions to the NEA International FEP List, NEA/RWM/R(2013)8, OECD NEA, Paris, September 2012 (published June 2014). 32

33 Session 3c ILW 03d 07 / ID 184. Disposal of High Level Waste SITEX, THE EUROPEAN NETWORK OF TECHNICAL EXPERTISE ORGANISATIONS FOR GEOLOGICAL DISPOSAL D. Pellegrini 1, F. Bernier 2, V. Detilleux 3, G. Hériard Dubreuil 4, A. Narkuniene 5, J. Miksova 6, M. Rocher 1 1 Radiation Protection and Nuclear Safety Institute (IRSN), France 2 Federal Agency for Nuclear Control (FANC), Belgium 3 Bel V, Belgium 4 MUTADIS, France 5 Lithuanian Energy Institute (LEI), Lithuania 6 Research Centre Rez (CV-REZ), Czech Republic contact of main author: delphine.pellegrini@irsn.fr Abstract. A European SITEX network is being prepared to ensure a sustainable capability for developing and coordinating joint and harmonized activities related to the independent Expertise Function in the field of deep geological disposal safety. Two successive EURATOM projects devoted to the preparation of this network worked on the needed set of activities, which entails strengthening the review of safety cases, developing a research strategy, interacting with civil society and training. This paper presents the main outlines of the ongoing second project called SITEX-II, with a specific focus on the Strategic Research Agenda issued recently. Key Words: Expertise Function Network, Geological Disposal, Safety Case Review, Civil Society Involvement. 1. Introduction The European Council Directive 2011/70/EURATOM of 19 July 2011 establishes a Community framework for the responsible and safe management of spent fuel and radioactive waste. In line with this Directive and in consistency with international high level safety standards issued by IAEA and WENRA, waste management organisations (WMOs) are developing a safety case for presenting the technical and organisational arguments that support the development of the national geological repository in each concerned country. As safety cases develop, the safety case review by regulatory bodies in the framework of the decision making process develops as well. In that context, organisations in charge of reviewing the safety case must in particular evaluate whether the elements of safety, and in particular that supported by scientific and technological results, are sufficiently convincing to be accepted by the regulatory authority as a basis for proceeding with the decision making process. In that context, there is a need at the international level for developing and coordinating activities associated with the regulatory review process of deep geological disposal. In 2012, the EURATOM FP7 SITEX ( Sustainable network for Independent Technical EXpertise of radioactive waste disposal ) project was launched in order to complement existing initiatives (ENSREG, WENRA, NEA/RWMC/Regulator Forum ) with the view to characterize the Expertise Function (see Figure 1) devoted to the technical review of a safety case at national level for deep geological disposal of radioactive waste. 33

34 FIG. 1. The Expertise Function and its interactions [1]. The SITEX-II Project ( ), a EURATOM Horizon2020 Coordination and Support Action, is aimed at practical implementation of the activities defined by the former SITEX project using the interaction modes developed by that project and with a view to further prepare the future Expertise Function network. SITEX-II brings together, as partners, representatives from 18 organisations involving National Regulatory Authorities (NRAs), Technical Support Organisations (TSOs), Research Entities (REs), Non-Governmental Organisations (NGOs), specialists in risk governance and an education institute, and involves interactions with a wider group of Civil Society (CS) participants. Its tasks include programming R&D, developing a joint review framework, training and tutoring for reviewing safety cases and interacting with CS, as detailed below, together with preparing an Action Plan that will set out the content and practical modalities of the future Expertise Function network. 2. Programming R&D The 2011/70/EURATOM Directive requires the Expertise Function to carry out its own horizontal and R&D activities, so that it is not dependent on those developed by the Implementer Function to make its own judgement. It is also stressed in IAEA safety guides that the Regulatory Body, and thus its supporting organisations (see Figure 1), may need to conduct or commission R&D in support of regulatory decisions (see IAEA GS-G-1.1 [2] (see 3.33) and IAEA GS-G-1.2 [3] (see 3.68)). SITEX-II therefore includes a task which pursues the general objective of further defining the Expertise Function s R&D programme necessary to ensure independent scientific and technical capabilities for reviewing a safety case for geological disposal. In this perspective, a Strategic Research Agenda (SRA) [4] has been issued, which will be completed by the Terms of Reference (ToR) for its implementation. This SRA has been also an input to the JOPRAD project (EURATOM Horizon2020 Coordination and Support Action Towards a Joint Programming Project on Radioactive Waste Disposal, ), which aim is to assess the feasibility and, if appropriate, to generate a proposal for Joint Programming activities that could be developed by WMOs, TSOs and/or REs in the field of Radioactive Waste Management (RWM), in particular geological disposal. The commitments of SITEX-II members for the development of the Expertise Function SRA are the following: the SRA is developed by applying a transparent methodology; the SRA addresses the needs associated with the different states of advancement of geological disposal programmes; 34

35 Session 3c ILW the concerns of CS participants are duly taken into consideration. The current version of the SRA entails topics relevant to the Expertise Function to assess whether geological disposal facilities are developed and will be constructed, operated and closed in a safe manner, for which a sufficient level of common interest has been expressed amongst the SITEX-II members. So, seven main topics related to pre and post-closure safety are considered in the SRA (1. Waste inventory, 2 Transient THMBC conditions in the nearfield, 3. Evolution of EBS material properties, 4. Radionuclide behaviour in disturbed EBS and HR, 5. Safety relevant operational aspects, 6. Managing uncertainties and the safety assessment, 7. Lifecycle of a disposal programme and its safety case). In addition to R&D activities, the needs for knowledge transfer (e.g. training, tutoring), for developing states of the art and for exchanging on practices and developing common positions are also identified. One particularly innovative development of the SRA relates to the introduction in the main topic n 7 of several holistic (complex) topics, for which both technical and societal aspects need to be investigated in an integrated manner, using specific interdisciplinary methodologies and involving CS participation. Also, regarding the other main topics, that are mainly technical, it came out essential to embed CS participation through the involvement of trained individuals, therefore offering the public the possibility to follow the development of this technical research, and to perform Knowledge Sharing and Interpretation (KSI) activities along the development of R&D results Figure 2 illustrates the associated issues and activities of common interest for Main Topic 1. SRA Main Topic and associated issues Main Topic 1: Waste inventory and source term #1. Uncertainty about databases and methodologies used for defining waste inventories(including historical waste) #2. Evolution of the waste inventory due to possible neutron activation #3. Understanding of the release processes and speciation of the radionuclides for spent fuel, vitrified and cemented waste #4. Waste acceptance criteria Research activities (experiment and/or modeling works) Exchange on practices, develop common positions Horizontal activities Develop states of the art FIG. 2. Associated issues and activities for Main Topic 1 of the Expertise Function SRA. 3. Developing a joint review framework Transfer of knowledge (eg. training) High-level safety requirements and regulatory expectations for the safety case at different phases of geological disposal facility development (conceptualization, siting, reference design, construction, operation, post-closure) are addressed by the EU Directive and international standards and recommendations (IAEA, ICRP, WENRA, etc.). This leads to a key objective for a second SITEX-II task to further develop a common understanding of the interpretation and proper implementation of safety requirements in the safety case for the six phases named above of geological disposal facility development. Position paper on the selected topics and technical guides related to the review of a safety case will be elaborated, accounting for existing initiatives and building upon return of experience at the international level. To date, SITEX-II participants have exchanged their views and experience on how to implement in practice the requirements and expectations related to optimisation of protection and to waste acceptance criteria. The next topics will be operational issues in regards with post-closure safety, with an introductory presentation by a GEOSAF2 representative, and programme for site characterization. 35

36 4. Training and tutoring for reviewing the Safety Case The third task of the ongoing project aims to implement a practical demonstration of training services that may be provided by the foreseen SITEX network. The training will be undertaken within the existing institute for expert training in nuclear safety (the European Nuclear Safety Training and Tutoring Institute, ENSTTI). The development of a training module at a generalist level with emphasis on the technical review of the safety case is ongoing. The module will be presented and evaluated in the pilot training session in Interactions with Civil Society The quality of the decision-making process, and its compliance to international rules and conventions, includes several requirements such as maintaining over time consultation and interaction with interested parties in the decision-making process. It is therefore crucial for the consistency of the SITEX-II project that interaction with CS is embedded all along the development of the future SITEX network. This is expected to contribute to transparency in the specific area of expertise, supporting the development of interactions between Expertise and Society functions at different levels of governance and at different steps of the decisionmaking process. SITEX-II involves, as partners, representatives of NGOs and interacts with CS participants through workshops covering three thematic tasks, namely: R&D, safety culture/review and governance. The results will be integrated in one deliverable addressing the conditions and means for developing interactions with CS in the framework of the foreseen SITEX network. The constructive discussions that took place to date within SITEX- II allowed both institutional and CS participants to exchange and challenge their views, fostering mutual understanding, notably through the elaboration of the SRA. The need for building mutual understanding led to the development of and was in turn dynamized by an innovative multi-stakeholders evaluation process and tool, allowing for a participative and comparative discussion of alternative scenarios of long-term RWM which target passive safety as their end point. 6. Conclusion The progress to date of the EURATOM Horizon2020 SITEX-II project shows that developing and coordinating joint and harmonized activities at the international level supporting the independent Expertise Function is achievable and promising in the field of geological disposal safety. Particularly, the involvement of of NGO representatives linked with a wider group of CS participants within SITEX-II should allow the Expertise Function to better account for societal concerns in its future networking activities, thus strengthening the decision making process. At this stage, the launching of the Expertise Function network is foreseen in , together with the Joint Programming if supported, meaning an ambitious calendar that calls for interested organizations that are not part of the SITEX-II project to express their interest now. This project has received funding from the EURATOM research and training programme under grant agreement No References [1] EC FP7 SITEX project, D6.1 Conditions for establishing a sustainable expertise network, [2] IAEA, Safety guide GS-G-1.1-Organization and staffing of the regulatory body for nuclear facilities, Vienna,

37 Session 3c ILW [3] IAEA, Safety guide GS-G-1.2-Review and assessment of nuclear facilities by the regulatory body, Vienna, [4] DETILLEUX, V. et al., Overview of the Strategic Research Agenda in the field of geological disposal of radioactive waste developed by the Expertise Function in the EC- H2020-SITEX-II project, EUROSAFE Forum 2016, 7th & 8th November 2016, Munich, Germany (in preparation). 37

38 03d 08 / ID 207. Disposal of High Level Waste RESEARCH AND DEVELOPMENT NEEDS IN A STEP-WISE PROCESS FOR THE NUCLEAR WASTE PROGRAMME IN SWEDEN A. Ström 1, K. Pers 2, J. Andersson 1, E. Ekeroth 1, A. Hedin 1 1 Swedish Nuclear Fuel and Waste Mgmt. Co. (SKB), Stockholm, Sweden 2 SKB International AB, Stockholm, Sweden contact of main author: anders.strom@skb.se Abstract The license holders have formed Swedish Nuclear Fuel and Waste Management Co. (SKB) to on their behalf develop and manage a programme for the research and development activities needed to manage and dispose of nuclear waste and spent nuclear fuel in a safe manner. The disposal of waste from decommissioning and dismantling of nuclear power plants is also part of SKB s assignment. Such a programme (RD&D Programme) has since 1986 been submitted every third year to the Swedish Radiation Safety Authority (SSM) for review as preparation for a Government decision on the programme. After more than 30 years of research and development regarding final disposal of spent nuclear fuel, an application under the Nuclear Activities Act for final disposal of spent nuclear fuel and an application under the Environmental Code for the KBS-3 system was submitted in March These license applications provided a summary of the current status of the development of the KBS-3-system and included a safety assessment. An application to extend the existing final repository for short-lived radioactive waste was submitted in The licensing processes are under way for both these repositories. Even though a large number of issues may be considered resolved regarding the systems there are still substantial technology development and demonstration efforts planned before disposal can begin and the facilities be operated as an industrial enterprise. Furthermore, SSM's regulations specify that development and licensing of nuclear facilities will be achieved through a stepwise process in which the requirements of the facility, its design and technical solutions is gradually established based on research, technology development and evaluation of safety after closure. RD&D programme 2016 was submitted in September. Adjusted to the current situation, needs for future research and technology development is based on the stepwise decision process described above. The milestones that are linked to major decision steps for new and extended facilities determine the required level of knowledge and development of technology. The safety reports together with the comments made by SSM in connection with the review of the applications, as well as audits of previous RD&D programmes, are the basis for the programme. Key Words: Waste disposal, spent fuel, research, technology development, safety analysis 1. Background and introduction The Swedish power industry has been generating electricity by means of nuclear power for more than 40 years. During this time, a large part of the system for management and disposal of the radioactive waste and the spent nuclear fuel has been built up. The system consists of the interim storage facility for spent nuclear fuel (Clab), the final repository for short-lived radioactive waste (SFR) and a system for transportation of nuclear waste. What remains to be done is to build and commission the system of facilities, the KBS-3 system, needed for final disposal of spent fuel shown on Figure 1. This work includes building a facility part for encapsulation of the spent nuclear fuel, developing transport casks for shipping canisters, and building a final repository where the canisters will be deposited. For disposal of short-lived low- and intermediate-level waste, the existing repository SFR will be extended, containers will be developed for transportation of long-lived waste, and 38

39 Session 3c ILW eventually a final repository for long-lived waste will also be built. The process of construction and commissioning a new facility, or extending an existing facility, consists of several phases. In 2011, after an extensive siting phase, SKB submitted an application under the Nuclear Activities Act for final disposal of spent nuclear fuel and an application under the Environmental Code for the KBS-3 system (encapsulated spent nuclear fuel in copper canisters with an insert of cast iron, embedded in bentonite clay at 500m depth in crystalline bedrock) adjacent to the Forsmark nuclear power plant site. In order to be able to dispose of all additional short-lived operational waste from dismantling, SKB submitted an application in 2014 for the extension of the SFR facility. FIG. 1. The Swedish system for management of nuclear waste. Under the Nuclear Activities Act, the nuclear power companies shall draw up a programme for the research and development activities and other measures needed to manage and dispose of the nuclear waste and the spent nuclear fuel in a safe manner and to decommission the nuclear power plants. The license holders have formed SKB to on their behalf develop and manage such a programme. The RD&D Programme [1, 2] has since 1986 been submitted every third year to the regulator SSM for review as preparation for a Government decision on the programme. This process for regular reporting and review of results and plans has contributed significantly to the development of a high scientific quality of the work and an open and transparent review mechanism. The regular review of the RD&D-programmes every third year has had a significant influence on the programme. 2. Implementation plan The implementation plan describes the measures needed to meet SKB s obligations and when applications and other legally mandated reports for the facilities are planned to be submitted. During 2015, decisions have been taken on an early shutdown of four reactors in Sweden. This means that the total amount of fuel that will be managed within the programme decreases. The remaining six reactors are planned to be operated for 60 years. Assuming that all the reactors have been taken out of service by 2045, SKB s three final repositories (the Nuclear Fuel Repository, SFR and SFL) can be closed in about 60 years. These times are important premises in the planning. The estimated start of construction for the Nuclear Fuel Repository is 2020 and that for the Encapsulation part of Clink is These facilities will be in operation simultaneously in 39

40 2030. The Encapsulation Project includes planning, design, construction and commissioning of the integrated facility for interim storage and encapsulation in Oskarshamn. For the Nuclear Fuel Repository Project the final phase of system design of the final repository s facility parts and technical systems has recently been completed. The extended repository for low and intermediate-level waste, SFR, is expected to be ready for operation in 2028 to meet the needs of the nuclear power industry to dispose of nuclear waste from operation and decommissioning of the nuclear power reactors. SKB plans to apply for a licence to build the next final repository for long lived waste, SFL, in around The licensing processes are under way and are expected to take several years. The relatively long time horizon covered by SKB s planning means that the planning premises may change in the meantime and be handled accordingly. Work is under way with safety analysis reports for the facilities which have to be submitted prior to the start of construction. This work is based on experience from the preparation of the safety analysis reports submitted with the applications, and from the reviews conducted within the on-going licensing processes. The construction projects and the work with safety analysis reports are primary beneficiaries of the technology development and scientific research that is being carried out. 3. Continued research and technology development For establishing nuclear facilities, planning is based on the stepwise decision process in the Nuclear Activities Act and SSM regulations. The safety analysis report (SAR) is central and should provide an overall view of how the safety of the facility is arranged in order to protect human health and the environment against nuclear accidents. The report shall reflect the facility as built, analysed and verified, as well as show how the requirements on its design, function, organisation and activities are met. The implementer needs to provide successively refined safety reports to the regulator. The planning and milestones related to decision steps in the form of applications and safety analysis reports determine when knowledge and development of the technology needs to have reached a certain level, while SSM s approval determines when SKB can commence construction and operation of the facilities. SKB has, as part of the applications of new and extended facilities now in progress, produced collective accounts of the state of knowledge and the status of technology development. In conjunction with this, the importance of remaining uncertainties regarding the ability to fulfil the requirements on protection of humans and the environment against radiation after closure of the repository has been evaluated. These reports, together with the viewpoints that have been submitted in conjunction with the license review and reviews of previous RD&D programmes, form a basis for the planned activities for research and technology development within various disciplines. The need for research and development activities can be divided into three main categories: The need for an increased process understanding, i.e. scientific understanding of the processes that influence the final disposal and thus the basis for assessing their importance to safety after closure. 40

41 Session 3c ILW The need for knowledge and competence regarding design, construction, manufacture and installation of the components included in the system. The need for knowledge and competence of inspection and testing to verify that the barriers and components are produced and installed according to approved specifications and thereby satisfy the requirements. Based on this, the research and technology development needed to solve the design and construction issues relating to the repositories, and the research needed to carry out assessments of the safety of the repositories post-closure, has been identified and justified. 4. Examples of important issues - research and technology development The comprehensive research, development and planning work conducted over four decades has led to many issues of importance for the nuclear waste programme being treated and resolved. Here, very brief accounts of the need for the research and development being identified for the remaining parts of the nuclear waste programme are exemplified. For the KBS-3 method copper canister is the containment barrier. Continued work concerns both the research on copper canister properties in the repository environment and technological development in order to be able to produce canisters, verify them against stipulated requirements and handle them in the KBS-3 system. For the assessment of postclosure safety, there are issues regarding corrosion and creep that require further research. Sulphide is the dominant long-term copper corroding agent in a KBS-3 repository. A better understanding of the details concerning sulphide corrosion strengthens the scientific basis for the safety assessment. The understanding of copper creep in the presence of mechanical loads is incomplete. In order to be able to improve the modelling of creep in the assessment of canister strength, among other things the understanding of how admixture of phosphorus leads to favourable creep properties needs to be strengthened. Clay materials are used in all three repositories: as buffer and backfill in the KBS-3 repository, in silo filling in SFR and as a barrier in the rock vault for the legacy waste in SFL. For the KBS-3 the design of the buffer, backfill and closure needs to be further developed prior to the continued design of the final repository as well as the production system for bentonite components. The need for measures for quality assurance during manufacturing, handling and installation needs to be further detailed. 5. Concluding remarks It is of utmost importance to address the issues within research and technology development that are most relevant for the development of new facilities at the time when they are needed and in a cost efficient way. The SKB RD&D Programme 2016 includes an up-to-date planning and presents those issues in a structured and step wise procedure based on the milestones for all new facilities and for other measures needed. REFERENCES [1] SKB. RD&D Programme Programme for research, development and demonstration of methods for the management and disposal of nuclear waste, SKB TR-13-18, Svensk Kärnbränslehantering AB, Sweden, 41

42 [2] SKB. RD&D Programme Programme for research, development and demonstration of methods for the management and disposal of nuclear waste, SKB TR-16-NN, Svensk Kärnbränslehantering AB, Sweden, translation in progress. 42

43 03d 09 / ID 26. Disposal of High Level Waste GERMANYS NEW ROUTE TOWARDS A REPOSITORY FOR HLW SCIENTIFIC CHALLENGES F. Charlier, B. Thomauske RWTH Aachen University, Institute of Nuclear Engineering and Technology Transfer (NET), Aachen, Germany contact of main author: charlier@net.rwth-aachen.de Abstract. Since 2011 Germany is pursuing a phase out strategy concerning the use of nuclear power for electricity production. This decision was influenced by the Fukushima event. In 2013 the federal government announced that they also had achieved an agreement with the Federal States in Germany on a law to restart the site selection for a repository for spent fuel and high active heat producing waste from scratch. The consequence of this law is a delay of at least two decades to start operation of a final disposal site. At first a commission had been installed to evaluate the Site Selection Law and to develop basic principles for site selection, including safety requirements and selection criteria for rock formations. The site selection then might start after the next federal election in 2017 at the earliest probably based on a new site selection law. A new repository site should be determined till 2031 and for this site the more detailed site investigation will take place followed by a detailed safety analysis, before the erection of the repository can start. Based on the present procedural steps, it seems to be rather unlikely to determine a repository site till There will be a delay of at least 20 years compared to the schedule given in the site selection law until a repository site can be determined. Therefore it is important to think about possibilities to accelerate the process without any reduction in safety. This paper presents main future needs for research and development on the German path towards a repository site for HLW. 1. Final Disposal of Radioactive Waste in Germany From 1979 until 2013 the salt dome of Gorleben was investigated for the disposal of high active heat generating waste. This site investigation was stopped in 2013 after a new site selection act came into power. This site selection act was evaluated by a commission. It is now intended to start a new site selection procedure from scratch including salt, clay and crystalline as host rocks. Besides for negligible heat generating waste the iron ore mine Konrad had been licensed in Since then it is transformed into a repository. It is expected that Konrad will start in operation around An overview over the German disposal situation is given in TABLE I. 43

44 TABLE I: Disposal Projects in Germany Project Geological Formation Purpose Actual Status Waste Gorleben Salt dome Repository for all types of radioactive waste especially high-level and heatgenerating waste All investigations are stopped in 2013 But will take part in the new site selection 17,000 t HLW/spent fuel New site selection Salt Clay Repository for highlevel and heatgenerating waste Evaluation of the site selection act 17,000 t HLW/spent fuel Crystalline Konrad since 1982 Iron ore Repository for long lived waste with negligible heat generation Licence issued 2002 Start of operation ,000 m 3 LLW/ILW Operation: 35 years 2. Site Selection Process The procedural steps to determine a repository site are [1]: 1. A first stage to evaluate the legal regulations and to determine general criteria. 2. Investigation of potential siting regions. 3. Exploration from above ground. 4. Exploration of the underground area. 5. Comparison of sites. 6. Recommendation of one site. 7. Determination of a site by federal law. 8. Licensing procedure for the proof of safety at the defined site based on a detailed underground exploration. 9. Construction of the facility after legal verification of the approval decision, if applicable. This stepwise approach - including the underground exploration - is based on the German final disposal concept from earlier times. At first starting from a white German map exclusion criteria will be applied. For the remaining areas, minimum criteria and weighing criteria will be adopted and result in regions or sites which may be suitable. Among these using safety analyses several regions will be selected which turn out to be the most suitable candidates for a site investigation from above ground. Based on the results of the site investigations from above ground, a few sites will be identified as the candidates with the highest expectations with respect to suitability. 44

45 After a site investigation of the host rock from below ground one site will be selected after safety analyses and proposed to become the site for which the licensing procedure should be performed. The site selection process leads to one site for which the licensing procedure will be initiated. The target of the site selection process is to find in a transparent way criteria based one site which is expected to then be the best possible solution. If it would turn out within the licensing process that the selected site cannot be licensed due to safety reasons based on new findings a setback has to be initiated and one has to go back one or two steps in the process depending on the new insights. 3. Paths Forward The commission has analysed the different potential solutions to dispose of high active heat generating nuclear waste. The preferred solutions called path - based on the present state of the art is the final disposal in deep geological formations in a mine [1]. Besides there are other potential solutions, where the technologies are not yet available but which may turn out as possible technologies for the treatment or disposal of theses waste stream. They should be analysed repeatedly after certain time steps. Especially the final disposal in deep boreholes offers an alternative to the disposal in a mine. But at the moment questions like recoverability or what if the disposal process fails are not yet answered. Here it is intended to watch the technology development. 4. Criteria The commission discussed geological and societal criteria but agreed in the main principle that safety has priority. All other criteria are seen as secondary with regard to this main important criterion. The criteria are differentiated between [1]: 45 Exclusion criteria Minimum requirements Weighing criteria The main principles for the site selection process are: Safety is of priority; Recoverability, reversibility; Step by step approach; No right of veto of the regions/sites but they should have the possibility that the process have to be iterated by one step; Transparency, Public participation and Stakeholder involvement. 5. R&D needs and scientific challenges The German Entsorgungskommissin (ESK) [2] describes the R&D demand as follows: 1. Specific R&D on host rocks: - Clay, Crystalline, Salt 2. R&D, independent of host rocks

46 - Safety concepts - Repository concepts - Interaction between repository, barriers and host rocks - Site evaluation - Characterization and comparison of sites - Retrievability and recoverability - Safety analyses and concepts for long-term safety In addition, R&D in the following topics is necessary: [The list should not be seen as being complete.] - Fleshing out the steps of the site selection process, - Development of canister requirements - New or further development of canister concepts - Application of the isolating rock zone -concept - Development and demonstration of handling the waste / the canisters - Development of concepts for following the site selection path transparent - Development of concepts for public participation - Development of concepts for stakeholder involvement 6. Conclusion The stepwise approach for finding a suitable site for a repository for Germany s heat generating waste will start in The preferred solution based on the present state of the art is the final disposal in deep geological formations in a mine. The scientific challenges and the upcoming R&D program are complex as shown in chapter 5 and due to the purposed step-back-option (learning process), the R&D program has to react flexible to new findings, demands or modifications of criteria or requirements. All R&D has to be identified and to be prioritized in the now starting site selection process. REFERENCES [1] Kommission Lagerung hoch radioaktiver Abfallstoffe, Abschlussbericht der Kommission Lagerung hoch radioaktiver Abfallstoffe, K.-Drs. 268 (2016). [2] Endlagerkommission (ESK), Stellungnahme der Entsorgungskommission, Endlagerforschung in Deutschland: Anmerkungen zu Forschungsinhalten und Forschungssteuerung, ESK (2016). 46

47 03d 10 / ID 32. Disposal of High Level Waste RECENT SAFETY ASSESSMENT OF A REFERENCE GEOLOGICAL DISPOSAL SYSTEM FOR RADIOACTIVE WASTE FROM PYRO-PROCESSING IN KOREA J.-W. Kim, D.-K. Cho, J. Jeong, M.-H. Baik, K. Kim Korea Atomic Energy Research Institute (KAERI), Daejeon, Korea contact of main author: jw_kim@kaeri.re.kr Abstract. For a long-term safety assessment to be comprehensive, complex scenarios should be assessed systematically by combining various scenarios with aleatory uncertainty. A methodology for a risk-based safety assessment of complex scenarios considering the long-term complementary impacts on the disposal system has been newly suggested by KAERI. This new methodology was recently implemented in an upgraded version of KAERI s TSPA model (K-PAM). KAERI s current TSPA model contains many necessary abstractions and a limit in associating the key physical processes. As a further study, the TSPA model will be moved to the process model level by utilizing a high-performance computing system. Key Words: Complex scenario; Risk-based safety assessment; K-PAM. 1. Introduction Since 2007, the Korea Atomic Energy Research Institute (KAERI) has studied the geological disposal of radioactive waste generated from the pyro-processing of PWR spent nuclear fuel [1]. The study mainly includes the characterization of geological media, the design of a reference disposal system, and the overall safety assessment of the disposal system. The characterization of geological media at different scales has been mainly conducted at the KAERI Underground Research Tunnel (KURT) area, the host rock of which is granite. The conceptual design of the reference disposal system is basically based on the Swedish KBS-3 concept. For the safety assessment of a hypothetical disposal system, a total system performance assessment (TSPA) model was developed using GoldSim. For a long-term safety assessment to be comprehensive, complex scenarios should be assessed systematically by combining various scenarios with aleatory uncertainty. In this study, a methodology for a risk-based safety assessment of complex scenarios considering the long-term complementary impacts on the disposal system is presented and implemented in an upgraded version of KAERI s TSPA model (K-PAM). For an illustration, a statistical analysis of historical seismic events and well exploitation in Korea was utilized to generate a complex scenario for a riskbased safety assessment. 2. Reference Disposal System KAERI presented a preliminary conceptual design of a geological disposal system for the radioactive wastes generated from the pyro-processing of PWR spent nuclear fuel. The radioactive wastes were classified into two groups: (1) low & intermediate-level metal waste which consists of hull materials and support frames and (2) high-level ceramic waste which is vitrified molten salt from the electrowinning process. The metal wastes are emplaced in a storage canister (stainless steel) and then the storage canisters are packaged by polymer concrete, so-called metal waste disposal package (MWDP). MWDPs are supposed to be stacked up with buffer materials in the tunnel at 200 m depth. The ceramic wastes are emplaced in a storage canister (stainless steel) and then the storage canisters are packaged in a disposal canister which consists of an inner container for the structural stability and an outer 47

48 shell for corrosion resistance. The disposal canisters are supposed to be emplaced with buffer materials in the borehole at 500 m depth (FIG. 1). < Metal Waste > < Reference Disposal System > < Ceramic Waste > FIG. 1. Conceptual design of a geological disposal system presented by KAERI. 3. Risk-based Safety Assessment 3.1.K-PAM Methodology The risk-based safety assessment methodology consists of 5 steps as shown in FIG. 2. The external events include natural disruptive events, such as earthquake, etc., and human intrusion. In the 1st step, the properties of those events related to the performance of the disposal system are digitized and represented by probability density functions (PDFs). In the case of an earthquake, for example, the properties can be the event occurrence rate, magnitude, distance from the hypocenter, etc. The PDFs of each property have to be carefully determined based on the historical records, a statistical analysis, expert judgments, etc. The PDFs of each property are converted into cumulative density functions (CDFs) for a scenario combination. In the 2nd step, how the external events will affect the disposal system is defined and the complex scenario generation criteria are determined. The external events will discriminatorily affect each part of the disposal system, such as an engineered barrier system (EBS), natural barrier system (NBS), and the biosphere. The impacts on the disposal system are also dependent on the properties of the external events. Some impacts can be irreversible so that the influence continues during the period of assessment, and some impacts can be reversible so that the disrupted parts are recovered after some time, or the influence of some repeating impacts can be increased gradually. This process also has to be carefully conducted based on the analogical interpretations of the experimental results and the relevant field data. 48

49 In the 3rd step, a complex scenario is generated based on the criteria defined in the previous step. Monte-Carlo sampling method is utilized as random numbers are independently generated and converted into the occurrence times and/or the values of the properties using the predefined CDFs for each property of the external events. The types of impacts by the external events are then determined based on the criteria. As all impacts on the disposal system are arranged in the process of time, a complex scenario is finally completed. For every iteration, a new complex scenario is preliminarily generated through this step. In the 4th step, each complex scenario developed in the 3rd step is simulated using the user-defined TSPA model. As the results of the scenario assessments, the exposure dose rates to the representative person are computed for each scenario. Because the complex scenario was randomly generated based on the criteria and their probabilities, the resulting FIG. 2. Flowchart of risk-based safety exposure dose rates already involve the assessment. probability of the scenario. In other words, if an exposure dose rate is obtained often from the iterations of scenario assessments, it implies that the scenario related to the exposure dose rate has a high occurrence probability. After each iteration, the exposure dose rates are cumulatively averaged and converted into the total risk using a dose-to-risk conversion factor. As the number of iterations increases, the results will be statistically stabilized. If the difference between the risks calculated in each iteration is less than the user-defined convergence criteria, it is assumed that the number of iterations is sufficient to consider exhaustively all possible scenarios. In the final step which is a post-process step, the final risk, the occurrence probabilities of each scenario, and the complementary safety indicators are computed as ordered. Additionally, sensitivity analysis can also be conducted in this step. 3.2.K-PAM Modeling System The methodology above was numerically implemented in an upgraded version of KAERI s TSPA model (K-PAM). The overall computing steps in FIG. 2 are conducted using Matlab except the scenario assessment (4th step) which is conducted using GoldSim. That is, a Matlab-based overall computing system is equipped with a scenario assessment module which was developed using GoldSim (FIG. 3). The GoldSim-based safety assessment model explains the source term, radionuclide transport in the EBS and far-field host rock (NBS), and radionuclide transfers in the biosphere. The radionuclide transport in the EBS includes the radionuclide release from a MWDP (metal waste) or disposal canister (ceramic waste), diffusive transport through buffer material, sorption, precipitation, and radioactive decay in the EBS. In the far-field host rock, radionuclide transports through the fractured rock undergoing sorption, precipitation, matrix diffusion, and radioactive decay. 49

50 Dose (msv/yr) / Risk (/yr) Session 3d HLW FIG. 3. GoldSim-based TSPA model and an illustrative result of the risk-based safety assessment. 3.3.Illustration An illustrative result of the risk-based safety assessment is depicted in FIG. 3. In the illustration, two external events, earthquake and well intrusion, were considered in the complex scenario generation. From the results, the computation was successfully converged into less than 1% risk-change after about 400 iterations. The time-series of dose for each iteration are depicted with gray line, and the median dose and the risk are depicted with black and red lines, respectively. 4. Concluding Remarks A methodology and a modeling system (K-PAM) for a risk-based safety assessment of complex scenarios considering the long-term complementary impacts on the disposal system were developed in this study. The results reasonably confirm the efficiency and stability of the modeling system. From the risk-based safety assessment of complex scenarios, the reliability, safety and public confidence of the disposal system is expected to be convinced more efficiently. KAERI s current TSPA model contains many necessary abstractions and a limit in associating the key physical processes. As a further study, the TSPA model will be moved to the process model level by utilizing a high-performance computing system. REFERENCES [1] KOREA ATOMIC ENERGY RESEARCH INSTITUTE, Geological Disposal of Pyroprocessed Waste from PWR Spent Fuel in Korea, KAERI/TR-4525/2011, KAERI, Korea (2011) Time (yr) 50

51 03d 11 / ID 34. Disposal of High Level Waste ASSESSMENT OF DECAY HEAT IN PROCESS OF SPENT NUCLEAR FUEL DISPOSAL Y. Kovbasenko State Scientific and Technical Centre on Nuclear and Radiation Safety (SSTC NRS), Kyiv, Ukraine contact of main author: yp_kovbasenko@sstc.com.ua Abstract. Residual energy release of standard VVER-1000 spent fuel assemblies was calculated with the U.S. SCALE code package for a storage period of years. WESTINGHOUSE (USA) and TVEL (Russia) fuel assemblies operating currently in the reactor cores of Ukrainian NPPs were considered. The calculations are provided for average geometrical, material and operating parameters of fuel assemblies. Upon the results obtained, empirical relations based on the sum of two exponential functions are proposed. They describe well the dependence of residual power release in spent fuel on the storage time from 100 to 1000 years. For the case of final disposal of spent fuel in sealed containers in geological rock formations, it is assumed that thermal radiation may be the only mechanism for heat removal from spent fuel. Based on the balance of power release in spent fuel and thermal radiation power of the blackbody, the time required for interim storage of spent fuel assemblies was conservatively assessed so that fuel temperature in final disposal would not exceed limiting values ( С). In this case, conservative assessments of the minimal required time of interim storage are from 100 to 200 years. Key Words: spent fuel assemblies, residual energy release, storage of spent fuel. 1. Introduction In the final disposal of spent nuclear fuel, it is assumed that there will be no monitoring operations or, if any, they will be minimized. Hence, it is very important to assess correctly the influence of the processes that occur in fuel on its storage parameters. One of these processes is residual power release in spent fuel. Storage of spent nuclear fuel in sealed cavity in deep geological formations is commonly considered as the main option for its final disposal. In this case, heat removal from fuel will be very limited and even insignificant residual power release can lead to substantial fuel heatup in the storage process. The final disposal is preceded by two stages: cooling a spent fuel in reactor pool and interim (often dry) storage. Their objective is to decrease residual heat release to an acceptable level. This paper provides preliminary assessments of the time required for cooling of spent fuel prior to its final disposal. 6. Determination of residual heat of spent fuel assemblies Consider the residual heat release in typical fuel of Ukrainian VVER-1000 NPPs produced by the Russian TVEL and U.S. Westinghouse fuel companies. To determine the residual heat of spent VVER-1000 fuel assemblies US SCALE code package was selected. The SCALE package includes computer modules, which combining programs and libraries to calculate one or another problem (criticality analysis, radiation safety, heat transfer, isotopic composition vs. burnup). The most complete description of the programs included in the SCALE is provided in [1]. The applicability of the SCALE code 51

52 package and its libraries of neutron-physical constants for modeling VVER fuel are considered in [2]. The calculations were performed with the use of standart 44GROUPNDF5 library of neutron-physical constants. Calculations were made for reactor cells of VVER-1000 fuel under the burnup level up to 50 GWt*day/tU in 4 year fuel cycle. This cells were composed of the typical modern fuel assemblies TVS-A of Russian TVEL suppliers (Fig.1) and new fuel assemblies FA-WR of Westinghouse company (Fig.2). The main features and differences in geometrical and material parameters of TVS-A and FA-WR used in the calculations are presented in Table I. The results of these calculations are shown in Fig. 3. The results demonstrate that residual heat in fuel assemblies of both types is quite close. For a period of years, the numerical values are described well by the following empirical dependence: P (Wt/t) = 424.4*(1.4*exp(-0.02*t+1.0)+0.6*exp(-0.003*t+0.15))+65, where t (years) is post-operational period. TABLE I: SOME DIFFERENCES IN GEOMETRY AND MATERIAL PARAMETERS OF TVS-A AND FA-WR Parameter TVS-А (TVEL) FA-WR (Westinghouse) Fuel stack length 3530 mm 3530 mm Central Zone length (nom.) 3530 mm mm Axial Blanket length (nom.) - 2 zone x mm Fuel mass (UO2), kg ± 5.0 Fuel pin (312 pieces) Enrichment, wt% 306*4.4%+ 6*3.6%(BA) 240*4.2%+60*3.9%+ 6*3.6%+6*3.0%(BA) 0.714% (blanket) Pellet ID / OD, mm 1.4 / / 7.84 Cladding ID / OD, mm 7.73 / / 9.14 Cladding material/ density, g/ccm alloy Э110 (an alloy of zirconium) / 6.45 alloy ZIRLOTM / 6.55 Central tube ID / OD, mm 11.0 / / 12.6 Material / density, g/ccm alloy Э635(an alloy of zirconium) / 6.45 alloy ZIRLOTM / 6.55 Guide tube (18 pieces) ID / OD, mm 10.9 / / 12.6 Material alloy Э635 alloy ZIRLOTM Spacer grid (13 pieces in fuel zone) Mass, g Material / density, g/ccm alloy Э110 / 6.45 alloy 718 / 8.18 Ribs (6 stiffener corners) Width / thickness, mm 52 / 0,65 - Material alloy Э635-52

53 Decay heat, W/t Session 3d HLW FIG.1. TVS-A model FIG.2. FA-WR model 900 ТВСА-4386 ТВС-WR382RR 424.4*(1.4*exp(-0.02*x+1.0)+0.6*exp(-0.003*x+0.15)) Time, year FIG.3. Decay heat vs. time 7. Determination of time required for interim storage of spent fuel In the final disposal of spent fuel in closed underground compartments, thermal radiation will be the main processes of heat removal from the fuel. If fuel assemblies are arranged in several layers and there is no good thermal contact between them, heat exchange between FAs will also mainly proceed through radiation. As a model to assess the amount of heat removed through radiation, use the well-known Stefan Boltzmann equation for a gray body: P= α σ (T 1 4 -T 2 4 ) S, where: 53

54 α - radiation coefficient (degree of blackness); σ = 5,67*10-8 W / (m 2 K 4 ) - the Stefan Boltzmann constant; T1 temperature of the emitting surface; T2 temperature of the compartment wall ; S area of the emitting surface. scheme 1, 1 FA scheme 2, 7 FA scheme 3, 19 FA scheme 4, 37 FA FIG.4. FA arrangement in final disposal According to the published data, the typical radiation coefficient of polished metals is α= [3, 4]. Using simple geometrical calculations, find S=0.47 m 2 area of one FA face. Temperature of the compartment wall is assumed to be Т 1 =50 о С=323 K. The limiting temperature of FA emitting surfaces in the storage process is accepted to be Т FAlimit = 300 о С = 573K. In accordance with the above results, FA power release will be Р~450 W after 50 years of cooling, Р~250 W after 100 years, Р~180 W after 150 years and Р~150 W after 200 years of cooling. If we assume that radiation comes from the surface of one FA 6 faces (configuration 1), then calculation with the Stefan Boltzmann equation for FA surface temperature gives: Р=σ(Т 2 4 -Т 1 4 )*6S Т 2 4 = Т 1 4 +Р/(6 α σ S)=(108,8 + 28,14)Е+8 Т 2 =349 К If we assume that radiation comes from the surface of 7 FAs 18 faces (configuration 2), then the calculation with the Stefan Boltzmann equation for surface temperature of the central FA gives: 7Р= α σ(т 3 4 -Т 2 4 )*18S Р= α σ(т 3 4 -Т 2 4 )*2,57 S Т 3 4 = Т 1 4 +Р/(6 α σ S)+Р/(2,57 α σ S)= (108,8 + 28, ,71)Е+8=202,65Е+8 (242,87Е+8) Т 3 =395К Continuing the calculations for dense packing of greater number of fuel assemblies, we obtain: TABLE II. FA TEMPERATURE Number FA / layers of FA 1/1 7/2 19/3 37/4 61/5 91/6 Central FA temperature after 50 years (α=0.7) 349 К 395 К 446 К 496 К 545 К 590 К Central FA temperature after 50 years (α=0.3) 377 К 453 К 528 К 597 К 661 К 720 К Central FA temperature after 100 years (α=0.3) 356 К 410 К 468 К 524 К 576 К 626 К Central FA temperature after 150 years (α=0.3) 348 К 391 К 440 К 490 К 537 К 582 К Central FA temperature after 200 years (α=0.3) 344 К 382 К 426 К 472 К 515 К 557 К 54

55 8. Conclusions Thus, for safe final disposal up to 19 fuel assemblies (3 layers) in an underground compartment 50 years is a sufficient period. For safe final disposal of 37 fuel assemblies (4 layers) in an underground compartment may require 100 years, of 61 fuel assemblies (5 layers) years and of 91 fuel assemblies may require 200 years of interim storage. Otherwise, the fuel assemblies may be overheated after sealing of the compartment and their integrity and configuration will be affected as a result. REFERENCES [1] SCALE User s Manual. NUREG/CR-0200 Revision 6. RNL/NUREG/CSD-2/V2/R6. [2] Y.Kovbasenko, V.Khalimonchuk, A.Kuchin, Y.Bilodid, M.Yeremenko, O.Dudka, NUREG/CR-6736, PNNL Validation of SCALE Sequence CSAS26 for Criticality Safety Analysis of VVER and RBMK Fuel Designs, Washington, U.S. NRC, [3] Neuer G., Thermal conductivity and thermal radiation properties of UO2, J. Non- Equilib. Thermodyn. 1, 3-23 (1976). [4] Siegel Robert, Howell John Thermal Radiation Heat Transfer, Fourth Edition, Taylor &Francis, NY,

56 03d 12 / ID 94. Disposal of High Level Waste ASSESSMENT OF PRE- AND POST-CLOSURE SAFETY IN THE NUMO SAFETY CASE FOR A GEOLOGICAL REPOSITORY S. Suzuki, K. Fujisaki, S. Kurosawa, K. Yamashina, A. Deguchi, H. Umeki Nuclear Waste Management Organization of Japan (NUMO), Tokyo, Japan contact of main author: ssuzuki@numo.or.jp Abstract. The NUMO safety case is established to improve the confidence of pre- and post-closure safety in the Japanese geological disposal programme at the current stage prior to selection of a site. The pre-closure safety case aims to assure both radiological and non-radiological protection of the public and workers. Radiological protection requires radiation shielding and radionuclide containment within the disposal facilities in case of operational perturbations. Operational perturbations, such as physical or thermal impacts on the waste-form, are analysed using an event tree method and possible, cost-effective counter-measures identified that would reduce their likelihood or mitigate their impact. Potential vulnerabilities of operational processes have been considered: most of these would pose little risk to the public, but the complexity of recovery operations and risks to workers could be significant. For protection from non-radiological hazards, the working environment will be maintained to ensure worker comfort and safety during normal operations. In many cases, requirements are set out in regulatory guidelines e.g. for the ventilation system. Further, underground tunnels and ventilation shafts should be laid out to facilitate ventilation pathways, taking transport routes for excavated rock and waste and required active / inactive zoning into consideration. Long-term, post-closure performance assessment is required to evaluate safety functions of specific repository systems, with consideration of uncertainties in a realistic and rational manner, excluding excess conservativeness. This is particularly required during site investigation to allow the pros and cons of potential sites to be identified and the appropriateness of particular repository concepts for such sites to be evaluated. Based on these requirements, an appropriate methodology has been developed for long-term performance assessment in this safety case. The methodology of scenario development, which results from a desire to combine a more conventional, bottom-up, FEP-based approach and a top-down method based on safety functions, is appropriate to this risk-informed assessment approach. This methodology, including overall procedures and associated toolkits, aims to increase traceability and transparency. Additionally, by clearly reflecting the purpose and context of the safety case and state-of-the-art knowledge, it assures appropriate degrees of completeness, comprehensiveness and sufficiency within the scenario development process. The methodology of safety analysis, which reflects the characteristics of site and repository design as faithfully as possible, has been improved. In particular, a radionuclide migration model for near-field scale ( several hundred meters) has been developed based on three-dimensional mass transport analysis that reflects key characteristics of the site and the associated repository design. Key Words: Geological disposal, vitrified waste, TRU wastes, safety case 1. Introduction NUMO has developed a safety case for co-disposal of HLW and TRU waste to reflect current boundary conditions in Japan, in particular siting based on an initial open call for communities to volunteer for initial site assessment. In particular, this involves addressing public concerns and actions by the Government to more strongly support moving forward with siting a geological repository, involving suggesting locations that are considered to be more scientifically suitable. The current Safety Case advances from the previous H12 Report [1], which formed the basis for establishing NUMO in 2000 as the implementing organisation and is considered the first generic safety case in Japan. The NUMO Safety Case has been developed to provide a 56

57 basic structure for subsequent safety cases that could be applied to any selected site, emphasising the practical approaches and methodology, which will be applicable for the conditions/constraints during an actual siting process. The NUMO Safety Case has been extended in key areas, including assessing extreme geological events during long-term repository evolution, widening discussion of both operational and post-closure safety, scenario development based on a risk-informed approach, etc. This paper describes the central issues of the safety case concerned with assessment of pre- and post-closure safety. 2. Assessment of pre-closure safety The reference inventory includes vitrified waste produced as a result of the reprocessing of spent fuel and TRU waste, which contains various types of intermediate level (but long lived) radioactive wastes produced by reprocessing and MOX fabrication. According to the final disposal plan [2], 40,000 packages of vitrified waste and a volume of 19,000 m 3 of TRU waste will be need to be disposed of. Radioactive protection of public and workers and nonradiological, conventional safety for workers during construction, operation and closure of repository are discussed Facility design for the radiological protection of the public and workers Radiological protection requires radiation shielding and radionuclide containment within the disposal facilities for all operations, extended to additionally cover potential operational perturbations. Radiation control and facility design are based on guidelines for other nuclear facilities [3]. Within radiation-controlled zones, most operations will be remote-handled or will involve appropriate shielding, avoiding any significant dose to workers. Under normal operations, radiological exposure of the public results only from highly penetrating radiation at or beyond the site boundary. Even assuming maximum exposure times, the expected dose beyond the boundary from the HLW handling facilities would be far below the upper limit of radiation exposure to the general public. To design safety measures, hazard scenarios were developed to identify operational perturbations resulting in physical or thermal impacts on the waste-form. The scenarios were made using event tree methodology and from this, possible, cost-effective counter-measures identified that would reduce their likelihood or mitigate their impact, on the basis of defensein-depth. TABLE 1 shows the multiple measures for the fire incident. TABLE 1: MITIGATION MEASURES FOR IDENTIFIED HAZARDS Level in event sequence diagram Measures Prevention of incident initiating fire Prevention of incidents providing ignition Elimination of combustible materials Prevention of fire propagation Elimination of combustible materials Detection of fire (e.g. Thermal/smoke detector) Fire extinguishing equipment Mitigation of radionuclide release Emergency exhaust filter system (if radionuclide accidents due to fire incident release is detected) Safety of workers (linked to Evacuation routes conventional safety issues) Emergency shelters Measures such as those mentioned above are designed to provide sufficient safety margins; however, the assessment conservatively assumes if all safety measures could fail. In practice, the mechanical robustness of metal packages effectively assures no release of radionuclides as a result of credible incidents in the underground facility. Potential vulnerabilities of 57

58 Connecting tunnel A Session 3d HLW operational processes have been considered: most of these would pose little risk to the public, but the complexity of recovery operations and risks to workers could be significant Facility design for the conventional safety of workers For non-radiological protection, the working environment will be maintained to ensure worker comfort and safety during normal operations. In many cases, requirements are set out in regulatory guidelines e.g. for the ventilation system. Further, underground tunnels and ventilation shafts should be laid out to facilitate ventilation pathways, taking transport routes for excavated rock and waste and required active / inactive zoning into consideration. For accident situations, such as a fire underground, the evacuation pathways would be routed along the air intake shaft, with emergency shelters provided at appropriate locations. To fulfill such requirements, we developed a simpler concept: involving a twin emplacement panel layout concept based on dead-end tunnels. In this concept, two horizontal connecting tunnels are utilized (FIGURE 1), with each tunnel operated independently for construction or operation. After finishing the construction of a disposal panel, the connecting tunnel and the constructed area are used for waste emplacement, while new panel excavation starts from the other connecting tunnel. Thus, the operation, ventilation and water drainage system will switch from normal area to a radiation-controlled area in a cyclic manner. This concept may also provide a simple evacuation pathway for emergencies such as fires. Mechanical Plug To access ramp Emplacement zone I (operational) Backfill completion tunnel Air intake Backfill intake Air Intake Construction material intake Emplacement Zone II (under construction) Disposal tunnel (TRU) Connecting tunnel (TRU) Exhaust shafts Backfilling direction Emplacement direction Excavation completion tunnel Excavation completion tunnel Ventilation air duct Excavation direction Bottom gallery Access ramp Exhaust shaft Intake shafts Connecting tunnel Emplacement zone <operational> <emplacement completed> Disposal tunnel <under construction> Emplacement zone III (plan) Exhaust Connecting tunnel B Exhaust Excavated rock removal Emplacement zone IV (plan) FIG. 1 Schematic view of the twin emplacement panel layout concept. 3. Assessment of post-closure safety 3.1. Framework for post-closure safety assessment Adopting a risk-informed assessment approach, assessment scenarios related to natural events and processes are classified into three categories related to the probability of their occurrence i.e. likely, less-likely and very unlikely. Scenarios related to human intrusion are treated based on a stylized approach, in line with the principle that such human intrusion scenarios are evaluated primarily to assess the robustness of the disposal system [4] Scenario development NUMO developed a hybrid scenario development methodology combining top-down (safety functions) and bottom-up (FEP-based) approaches in a complementary manner [5]. Specifically, the variables which influence a safety function allocated to a component of the system are defined, and the factors which influence these variables are selected from the FEP database (FIGURE 2). The treatment of each factor in a specific scenarios is determined by assessing the probability and significance of its occurrence. 58

59 Top Down State variable 1 State variable 2 Influencing factor 1-1 Influencing factor 2-1 Safety function of certain component Influencing factor 1-2 Influencing factor 2-2 Influencing factor 3-1 State variable 3 FEP Database FIG. 2 A fishbone diagram which shows the relationship between a safety function and influencing factors Modelling of radionuclide migration The safety analysis methodology has been improved to reflect the characteristics of site and repository design as faithfully as possible. In particular, a radionuclide migration model for the near-field scale ( several hundred metres) has been developed based on a threedimensional mass transport analysis that represents key characteristics of the site and the associated repository design. 3-D solute transport pathways are evaluated by a particle tracking method. The various calculation cases for the safety assessment scenarios should be carried out flexibly and efficiently, so radionuclide migration analysis taking account of retardation processes is conducted by using 1-D model. To better represent the case examined, the 1-D radionuclide model is fit to the solute transport properties obtained through 3-D particle tracking to create a 1-D multi-channel model. 4. Summary The pre- and post-closure safety cases were demonstrated. The R&D will be continued to improve the confidence in Japan throughout the siting and development of repository. REFERENCES [1] Japan Nuclear Cycle Development Institute, H12: Project to establish the scientific and technical basis for HLW disposal in Japan, JNC-TN , (2000). [2] Ministry of Economy, Trade and Industries: Policy of the Final Disposal of Designated Radioactive Waste (Cabinet Decision on May 22, 2015) (In Japanese), (2015). [3] Nuclear Regulation Authority, The new regulatory guideline for the HLW storage, (2013). [4] ICRP, Radiological Protection in Geological Disposal of Long-lived Solid Radioactive Waste, ICRP Publication 122, Ann. ICRP 42 (3), (2013). [5] Kurosawa, S., et al., Advances in scenario development for a deep geological repository in Japan, Proceedings of Global 2015, Paris, September (2015). 59

60 03d 13 / ID 96. Disposal of High Low Level Waste RESEARCH, DEVELOPMENT AND DEMONSTRATION PROJECTS AT THE JOSEF UNDERGROUND LABORATORY J. Stastka, J. Pacovsky Czech Technical University in Prague, Prague, Czech Republic contact of main author: jiri.stastka@fsv.cvut.cz Abstract. The Centre of Experimental Geotechnics (CEG), the department of the Faculty of Civil Engineering, Czech Technical University in Prague, is a full member of the Underground Research Facility Network for Geological Disposal (IAEA, URF Network). The main role of the URF Network is to establish a community for sharing experience and learning in the field of the geological disposal of radioactive waste. The CEG operates the Josef Underground Laboratory, and the extensive underground laboratory space available at this facility provides a unique background for experimental research, education, training and demonstration activities relating to the geological disposal of radioactive waste. Although the Josef facility is not intended for waste disposal, it does play a very important role in terms of the early stage of the site selection process for the Czech deep geological repository. The Expert Cooperation in the Construction of the first Czech Underground Migration Laboratory with the Potential Application of Active Tracers project makes up one of the most important research projects currently underway at the Josef facility. The objective is to obtain the knowledge from foreign partners necessary for putting the first in-situ underground laboratory with the potential application of active tracers into operation in the Czech Republic. As such research has not yet been conducted in the Czech Republic, it was essential to engage the involvement of foreign specialists. The participation of Swiss experts from the NAGRA organisation allows both the design and subsequent implementation of the experimental programme at the Josef facility s laboratories in such a way that the research and training processes are effective and so as to avoid the repetition of outdated experimental procedures and research topics. The ongoing Mock-up Josef experiment, which consists of an in-situ physical model simulating the vertical emplacement of a container with spent nuclear fuel, provides a further example of one of the more important experiments underway at the facility. The in-situ experiment involves research into the effects of heat and groundwater on the bentonite buffer surrounding a heater which simulates a spent nuclear fuel container emplaced in an underground repository. This article provides information on the Josef Underground Laboratory and its rich history of RD & D projects concerning the development of the Czech deep geological repository. Key Words: Josef Underground Laboratory, geological disposal, migration laboratory, Mock-up Josef 1. Introduction The Centre of Experimental Geotechnics (CEG) represents one of the most unique departments at the Faculty of Civil Engineering, CTU in Prague. In addition to providing teaching courses, mainly of a practical nature, in the field of geotechnics, it also specialises in the conducting of complex RD&D projects. One of the research facility s most important roles is to provide practical in-situ instruction in the fields of geotechnical engineering, geology, geochemistry, radiochemistry and radioecology. The training of future experts in this authentic underground setting also frequently involves the participation of other Czech universities and experienced specialists from outside the academic sphere. The IAEA (International Atomic Energy Agency) has added the CEG to its prestigious list of international training centres. In addition to teaching and training, the CEG is heavily involved in a wide range of research and development activities; indeed, the Josef Underground Laboratory, operated by the CEG, is currently being used for research purposes 60

61 in connection with a number of European Union-supported international experimental projects addressing a wide range of issues related to deep repository radioactive waste disposal (TIMODAZ - FP6, FORGE - FP7, PETRUS II - FP7, DOPAS, etc.) as well as several domestic projects (Mock-up Josef, etc.) supported by the Czech Ministry of Industry and Trade, the Czech Science Foundation and the Czech Radioactive Waste Repository Authority (SURAO). 2. Research, Development and Demonstration Projects The Josef Underground Laboratory offers more than 5km of a total of 8km of galleries (driven during the investigation of the Mokrsko Čelina gold deposits in the period ) for teaching and research purposes. No less than seven international and domestic research projects are currently underway at the Josef facility and a further four projects are in the preparation stage. Whilst the CEG is gradually extending the range of research and educational activities into other scientific fields, the main theme of both research projects and educational courses involves issues concerning the safe disposal of spent nuclear fuel in deep geological repositories including research into migration processes underway in real rock environments currently being conducted by the CEG in cooperation principally with the Nuclear Research Institute - Řež (ÚJV Řež), but also including a number of other partner institutions. Since the first section (Čelina West) of the Josef underground complex of galleries, with a total length of 650m (Fig. 1), was opened for educational and research purposes in 2007, the total length of reconstructed galleries has been gradually extended to over 5km (Čelina West, Čelina East and Mokrsko West). The granitic rock complex (the Mokrsko West section) has been equipped with core-forced air ventilation, a power distribution network, water supply systems and a high-speed internet optical cable network. The various core distribution systems installed at the facility will serve for the connection of the niche selected for the construction of the migration laboratory for research involving the application of active tracers. In 2010, the first student laboratory for teaching in the field of the disposal of hazardous substances and gases was constructed in the granitic rock medium section of the underground complex. In 2013, the CEG FCE CTU opened a migration laboratory for research involving non-active tracers in the Čelina West underground section (in the vicinity of the entry portals to the underground complex). The migration laboratory was built as part of the TA CR Determination of the Migration Parameters of Minerals with Fissure Permeability using Fluorescent Solutions project. Clearly, therefore, the CEG has extensive experience in terms of the reconstruction of underground galleries for the needs of specialised laboratories. In 2011, the CEG opened a new facility within the Josef complex the Josef Regional Underground Research Centre (Josef URC) which involved the complete reconstruction of a surface building to include an experimental hall, laboratories and other support facilities. This building provides the necessary backup services for the experimental research conducted in the underground complex. 61

62 FIG 1. Horizontal layout of the Josef Underground Laboratory (left) and pictures of the entrance to the underground complex; the upper-right picture shows the entrance to the underground complex prior to reconstruction in 2006; the lower-right picture shows the entrance portals today At the beginning of 2013, the CEG submitted the Inter University Laboratory for the In-situ Teaching of Transport Processes in a Real Rock Environment development project for approval to the Ministry of Education, Youth and Sport of the Czech Republic; the coresearcher consists of ICT Prague. In December 2013, together with a number of research partners (with ÚJV Řež as the senior researcher), the CEG applied to the Technology Agency of the Czech Republic (TA CR) to conduct the PAMIRE project (Transfer of Granitic Rock Parameters from the Micro Scale to the Real Rock Massif Scale). However, since neither of the above projects provide for financial support for the participation of foreign experts, the CEG subsequently decided to apply for a grant from the Partnership Fund of the Swiss-Czech cooperation programme. Consequently, a new project entitled Expert Cooperation in the Construction of the first Czech Underground Migration Laboratory with the Potential for the Application of Active Tracers commenced in Up to this time, research into migration processes in the real environment of an underground laboratory using active tracers was allowed at just two other European facilities SKB, Sweden and NAGRA, Switzerland. Importantly, the CEG (and ÚJV Řež) has enjoyed extensive cooperation with both these facilities in the past in the context of the EU FP6 and FP7 (Euratom) research projects. It is generally recognised that the Swiss organisation NAGRA, which runs the Grimsel Test Site (GTS) underground laboratory, employs the most experienced experts in the field of migration research. Therefore, the objective of the project, which will include consultation with Swiss experts, mutual visits and a series of bilateral workshops, is to obtain the knowledge necessary for putting into operation the Czech Republic s first in-situ underground laboratory with the potential for the application of active tracers. 62

63 3. Mock-up Josef Experiment Since the geological disposal of high-level radioactive waste is based on the multi-barrier concept, including the use of bentonite, the Centre of Experimental Geotechnics decided to construct the first Czech in-situ mock-up model of a disposal place employing a bentonite barrier. The experimental model, named Mock-up Josef, enjoys the active support of the Czech Radioactive Waste Repository Authority (SURAO). The project, which commenced in 2012, was planned to run for four years, i.e. to 2016 or up to the time the bentonite in the model reached full saturation. The physical model, which is situated in the Josef Underground Laboratory, is being loaded with underground water and features a heater which simulates the heat produced by the container with spent nuclear fuel enclosed by the bentonite layer. The model consists of a barrier made up of bentonite blocks, a heater, a comprehensive monitoring system and stainless steel construction equipment. The model was constructed in the Josef surface laboratory and subsequently transported to the selected niche in the Josef underground complex. The model was placed within a vertical disposal hole with a diameter of 750mm and a depth of 2500mm in December The experiment is located in the granitic section of the Josef facility (the Czech DGR development programme assumes that the future DGR will be constructed in granite host rock). 4. Conclusions Cooperation with international institutions provides an effective way in which to advance the Czech geological disposal of high-level radioactive waste programme. The complex research, development and demonstration projects conducted at the Josef facility provide information, experience and important data relating to the various components of the disposal system. 5. Acknowledgements Part of the work reported herein was supported by funding from Switzerland through the Swiss Contribution to the enlarged European Union. REFERENCES [1] PACOVSKÝ, J.; VAŠÍČEK, R.; Josef Regional Underground Research Centre - a New and Attractive Location for Interdisciplinary Teaching, Research and Training in the Field of Nuclear Engineering; In: Proceedings of the 17th Pacific Basin Nuclear Conference. (2010). ISBN [2] SVOBODA, J.; VAŠÍČEK, R.; The Josef UEF - a new location for in-situ physical modelling; In: ICPMG th International Conference on Physical Modelling in Geotechnics. (2010). ISBN [3] ŠŤÁSTKA, J.; Mock-up Josef Demonstration Experiment; In: Tunel, vol. 23, no. 2, pp , (2014). ISSN

64 03d 14 / ID 125. Disposal of High Level Waste THE MANAGEMENT OF USED (SPENT) FUEL AND HIGH LEVEL WASTE IN SOUTH AFRICA V Maree 1, A Carolissen 2 1 National Nuclear Regulator (NNR), Cape Town, South Africa 2 National Radioactive Waste Disposal Institute (NRWDI), Pretoria, South Africa contact of main author: vmaree@nnr.co.za Abstract. As a country with a nuclear power program and radioisotope production facility, the Republic of South Africa (RSA) generates Used Nuclear Fuel (UNF) and radioactive waste through numerous activities. The cornerstone of South Africa s approach to addressing radioactive waste management is the Radioactive Waste Management Policy and Strategy for the Republic of South Africa. The Policy and Strategy serves as a national commitment to address radioactive waste management in a coordinated and cooperative manner and represents a comprehensive radioactive waste governance framework by formulating, in addition to nuclear and other applicable legislation, a policy and implementation strategy developed in consultation with all stakeholders. In accordance with the Policy and Strategy, final disposal is regarded as the ultimate step in the radioactive waste management process, although a stepwise waste management process is acceptable. Long-term storage of specific types of waste, such as High-Level Waste (HLW), long-lived waste and high activity disused radioactive sources, may be regarded as one of the steps in the management process. This poster presents the South African National Radioactive Waste Management Model with a description of: the radioactive waste management governance framework; the current HLW and UNF management, the management option and UNF strategies. Also the poster addresses consideration of the lessons learnt from the Fukushima accident and its impact on future radioactive waste management strategies and options, plans related to possible long term operation of the existing nuclear power plants, introduction of new nuclear power plants and public acceptance and challenges from anti-nuclear groups. Key Words: Used Nuclear Fuel; High Level Waste; South Africa s Management Strategies; Challenges. 1. Introduction The Republic of South Africa (RSA) recognizes the importance of the safe management of spent fuel and radioactive waste, for this reason the country is a contracting party to the International Atomic Energy Agency (IAEA) Joint Convention on the Safety of Spent Nuclear Fuel Management and Safety of Radioactive Waste Management [1]. The Joint Convention provides for the establishment and maintenance of a legislative and regulatory framework to govern the safety of spent fuel and radioactive waste management. South Africa fulfills its obligations under the Joint Convention by the establishment of a Radioactive Waste Management Policy and Strategy for the Republic of South Africa (RWMP&S) [2] and has invited the IAEA to conduct the Integrated Nuclear Infrastructure Review (INIR) mission in The INIR mission has recommended that South Africa develop an integrated national Nuclear Fuel Cycle strategy, including Used Nuclear Fuel (UNF)/High Level Waste (HLW) disposal [3]. South Africa already has in place a strong national radioactive waste management model and is considering different options and strategies to address the long term management of the UNF and HLW as recommended by the INIR. 2. Background The past strategic programs and the current nuclear programs contribute to the generation of HLW and UNF. HLW for legal and regulatory purposes, is defined as waste with levels of 64

65 activity concentration high enough to generate significant quantities of heat (>2kW/m3), or waste with large amounts of long lived radionuclides. In 1991, South Africa signed the Nuclear Non-Proliferation Treaty and in 1993 voluntarily announced the dismantling of its nuclear weapons programme, HLW was generated. In the South African context, HLW doesn t include fuel coming from the irradiated fuel reactor cycle. The term used fuel is used instead of spent fuel because used fuel is considered to have useful material and is not classified as radioactive waste. UNF is produced in two main nuclear facilities: The South African Nuclear Energy Corporation (Necsa) and Koeberg Nuclear Power Station (KNPS). Necsa, located at Pelindaba 30 km west of Pretoria operates a 20 Megawatt tank-in-pool type nuclear research reactor: SAFARI-1 (Fig.2.). The research reactor has been in operation for 50 years and is used in the production of medical radioisotopes and nuclear research. KNPS is the only nuclear power plant in Africa and is comprised of 2 Framatome PWR reactors of 900 Mwe each operated by the State Own Company Eskom. KNPS is in operation since 1984 and situated on the Atlantic coast 40 kilometers north of Cape Town (Fig.2.). 3. South African Radioactive Waste Management Model The overarching objective of radioactive waste management is to deal with radioactive waste in a manner that protects human health and the environment now and in the future without imposing undue burdens on future generations. 3.1.Radioactive Waste Management Governance Framework The following diagram depicts the Governance Framework for radioactive waste in RSA: Fig.1. Schematic Governance Framework for Radioactive Waste Management. It is imperative to note that the legislative and regulatory framework for radioactive waste management and disposal is informed by and gives effect to: Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management [1]; National Radioactive Waste Disposal Institute Act (NRWDIA) [4]; Nuclear Energy Act [5]; RWMP&S. 65

66 In order to ensure that radioactive wastes are managed safely, the governance framework makes provision for: an established legislative and regulatory framework, the necessary organisations for implementation and providing oversight of waste management operations and facility development. Independence between the Regulator, Waste Generators and repository operator is the key to ensure that the RSA has an integrated and sustainable approach to ensure that the long term management is executed. RSA s approach to addressing radioactive waste management issues is RWMP&S which sets a general policy for dealing with all radioactive waste from the nuclear fuel cycle. RWMP&S developed in consultation with all stakeholders and transparency s principle serves as a national commitment to address radioactive waste management. RWMP&S puts forward the following hierarchy of waste management options to be followed, where practicable: avoiding waste and minimisation; reuse, reprocessing and recycling; storage; conditioning and final disposal. Final disposal is regarded as the ultimate step in the radioactive waste management process. The RWMP&S establishes the National Committee on Radioactive Waste Management (NCRWM). This committee is constituted by representatives from different organs of state. One of the committee responsibilities is to evaluate the radioactive waste plans submitted by radioactive waste generators and to provide recommendations to the Minister of Energy. The RWMP&S also makes provision for a National Radioactive Waste Management Fund managed by the South African Government to ensure sufficient provision for the long term management of radioactive waste with the principle that the Polluter pays. The NRWDIA became effective in December The NRWDIA endorsed the establishment of the National Radioactive Waste Disposal Institute (NRWDI) which is a national public entity. The Institute is mandated to discharge a Ministerial institutional obligation with respect to the management of radioactive waste disposal and related waste. The RSA acceded to the Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management (Joint Convention) in One of the objects of the NRWDI is to fulfil national obligations in respect of the long term management of radioactive waste disposal and related waste management activities as dictated by the Joint Convention. According to Section 5 of NRWDIA, the Institute must, inter alia, (a) perform any function that may be assigned to it by the Minister in terms of Section 55(2) of the Nuclear Energy Act, in relation to radioactive waste disposal; (e) manage, operate and monitor operational radioactive waste disposal facilities, including related storage and predisposal management of radioactive waste at disposal sites; (g) investigate the need for any new radioactive waste disposal facilities and site, design and construct such new facilities as may be required; (h) conduct research and develop plans for the long-term management of radioactive waste storage and disposal. 3.2.The Current HLW and UNF Management At present nuclear installations in South Africa use a combination of wet and dry storage for used nuclear fuel. UNF from the KNPS is currently stored in pools on the site as well as in casks designed and constructed for storage of used nuclear fuel. The used nuclear fuel from the SAFARI-1 Research Reactor is initially stored in the reactor pool for at least two years to facilitate cooling of the used fuel prior to it being transferred to an authorised dry storage facility on the Pelindaba site. Some HLW is stored on the same site. 66

67 3.3.The Management Option and Strategies for HLW and UNF The RWMS&P clearly indicates that storage on these sites is not sustainable in the long term and considers the following waste management options for UNF and HLW: long-term above ground storage on an off-site licensed facility; reprocessing, conditioning and recycling; direct deep geological disposal and transmutation. Regardless of any UNF/HLW management strategy chosen, a Centralized Interim long term off-site Storage Facility (CISF) and Deep Geological Repository (DGR) for final disposal will be required. Like any option chosen for the UNF, the DGR needs to be technically sound, socially acceptable, environmentally responsible and economically feasible. Transmutation requires major investment, two cycles options for the management of UNF can be considered: open cycle and closed cycle. Open cycle without recycling/ reprocessing: The fuel will be stored at the reactor site and will be transferred to a centralised off-site storage pending the final decision or, after the transfer to a centralised off-site storage, the UNF will be directly disposed. Closed cycle with recycling/ reprocessing: Firstly, the UNF will be stored at the reactor site for a specific period, secondly will be reprocessed and finally the UNF will be disposed or, after the transfer to a centralised off-site storage, the UNF will be reprocessed were the useful material will be reused and the waste will be disposed. Currently the solid low level and intermediate level waste from KNPS and Necsa are disposed of at the national radioactive waste disposal facility Vaalputs located in the Northern Cape Province (Fig.2.). Preliminary investigation in the early 90s has indicated that Vaalputs has suitable characteristics that would make this site a favourite candidate to host the CISF and a DGR. FIG. 2. Geographical location of selected nuclear facilities. The RSA has addressed the INIR mission s recommendation by drafting a new policy detailing options and strategies for UNF/ HLW management. The document was finalized and in currently under review by the Cabinet of the RSA. 4. Key Challenges One of the lessons learnt from the Fukushima accident was the importance to limit the UNF inventory on-site. On-site storage should only be for cooling purposes of the UNF. The RSA Government has committed to establish and operate a centralized off-site interim storage 67

68 facility by 2025 and a deep geological repository by 2065 [6]. In addition, centralized off-site interim storage facility will provide South Africa with the flexibility to make an informed decision with regard to fuel cycle strategy (open or closed) Provision must be made for additional waste storage/ disposal due to potential operation of new nuclear power plants as the RSA is considering a new nuclear build programme and the possible long term operation of the KNPS. The Fukushima accident has eroded the confidence of the public in nuclear power and safe radioactive waste management. Hence, to be successful the waste management programme must overcome this negative perception as technical competence is not enough to ensure and instill stakeholder trust and acceptability. Waste Generators in RSA must still submit their radioactive waste management plans for review and to the NCRWM. This committee will determine the funding strategy and requirements for sustainable long term operation of NRWDI. Funding is required for disposal activities, research and development including investigations into waste management/disposal options. The process of the site s selection for the centralized off-site interim storage facility and deep geological repository must be developed for licensing. 5. Conclusion South Africa has an integrated and extensive national radioactive waste management model which considers the different options and strategies to address the long term management of the UNF and HLW as recommended by the INIR. Final disposal is regarded as the ultimate step in the radioactive waste management process. In spite of challenges and irrespective of the fuel strategy chosen, it is inescapable that South Africa needs the following waste management infrastructure namely (i) CISF and (ii) a DGR. Finally, the RSA must develop and implement a comprehensive communication strategy and plan to demystify and decipher the public s fears regarding the management of radioactive waste and to deepen and strengthen stakeholder acceptance, confidence and trust. REFERENCES [1] Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management (IAEA, 1997). [2] Radioactive Waste Management Policy and Strategy for the Republic of South Africa (2005). [3] Department of Energy, Media Statement: Nuclear Procurement Process Update, 14 July [4] National Radioactive Waste Disposal Institute Act (NRWDIA), Act 53 of [5] Nuclear Energy Act, 1999, (Act No. 46 of 1999). [6] South African National Report on the Compliance to Obligations under the Joint Convention on Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management, September

69 03d 15 / ID 140. Disposal of High Level Waste REGULATORY EXPERIENCES IN REVIEWING CONSTRUCTION LICENSE APPLICATION FOR THE DISPOSAL OF SPENT NUCLEAR FUEL IN FINLAND J. Leino Radiation and Nuclear Safety Authority, STUK, Helsinki, Finland contact of main author: jaakko.leino@stuk.fi Abstract. Finland is one of the first countries in the world in developing a disposal solution for spent nuclear fuel (SNF). The Construction License Application (CLA) for the Olkiluoto SNF encapsulation and disposal facility was submitted by Posiva, the implementer, to the authorities at the end of 2012 and the Government granted construction license in November The post-closure safety case submitted as part of the CLA was reviewed during The CLA covered both operational safety (PSAR) and post-closure safety. In this paper, experiences gathered during the review process post-closure safety case are discussed. During the review process some practices proved to be good but the process revealed also some needs for improvements for the next licensing phase. Key Words: nuclear waste, spent nuclear fuel, disposal, review of post-closure safety case. 1. Introduction The safety case submitted as a part of the CLA was reviewed during The safety case covered both operational safety (PSAR) and post-closure safety documentation. The actual review process consisted of three phases: the initial phase, detailed review and assessment phase, and then the finalizing phase. The last phase consisted of preparation of conclusions, writing and finalization of statements, decisions and review reports. During the pre-licensing phase STUK started the work aiming to the readiness to review the construction license application. STUK also increased the own competence and resources to be prepared for the review. A strategic resource plan was made and the amount of people working mainly for the waste management was increased. STUK made also framework contract with 13 external experts to support STUK during the review of post-closure safety case. 2. Regulatory experiences during pre-licensing phase In the pre-licensing phase before the actual review phase STUK assessed Posiva s R&D work and draft documentation, planned the actual review process and increased its regulatory resources and competence. While assessing Posiva s R&D work and draft documentation STUK created a list of key safety concerns and had an active communication with Posiva. Active communication consisted of giving feedback to Posiva, discussions concerning the content and structure of the safety case, expressing regulatory expectations, hearing of Posiva s expectations and forming of common understanding of the regulatory requirements. The regulatory assessment of safety is, of course, done against regulatory safety requirements i.e. the government decree and regulatory guidance (STUK YVL guides). In this phase STUK s approach was initially safety issue oriented and bottom- up assessment. This was partly because of the regulatory safety requirements were not detailed enough. Thus, STUK started to develop a more structured review and assessment process for the safety case review 69

70 and prepared a so called review plan that would be a strong basis for the review and would guide the review. The review plan changed STUK s approach to more regulatory requirement oriented and safety related review basis. The review plan was seen necessary since it was acknowledged that addressing single technical concerns in many cases did not lead to better overall understanding of safety and often the linkage to safety was not very clear thus the review plan collects earlier regulatory observations and expectations for the safety case. Regulatory safety requirements given in the government decree on the safety of disposal of nuclear waste and regulatory guides were linked to these observations and expectations. The review plan was used as guidance both for internal and external experts participating in the review of the safety case. It also formed a basic structure for STUK s safety evaluation reports for operational and post-closure safety. 3. Regulatory experiences during initial review phase In the initial review phase STUK performed the completeness review of the safety case during the first three months. The aim of this phase was to verify that the safety case contained all main elements requested by the YVL guides. The conclusion from this completeness review was that Posiva had delivered most of the documentation required and STUK could continue the review despite some missing parts of the safety case. Nevertheless, STUK requested Posiva to deliver the missing parts of the documentation to STUK and to update some parts of the documentation which were considered to be too general. Based on the completeness review observations, STUK made several requests for additional information because of the adequacy and the quality of the safety case documentation. The extent and structure of the post-closure safety case documentation were such that it was challenging to check the adequacy and the quality of the elements during the initial review phase. However, this kind of completeness review proved to be useful tool to identify the completeness of the safety case. It also helped to identify any shortcomings and allowed STUK to continue its review and move on to the detailed review phase on areas that were found to be complete. 4. Regulatory experiences during detailed review phase The detailed review phase took place three to four months after the initial review phase depending on the documentation. The review plan which facilitated the review process of post closure safety documentation can be considered as a good practice. The review plan that STUK developed beforehand was seen as a strong basis for the review of CLA. Although, in the last phase of the review it was identified that even more detailed review guidance was needed in the next licensing phase. The more detailed guidance or review plan should consist of guidance on how to review and assess all the elements collected in the review plan, and elements of setting up the level of adequacy and also what are the most important and safety significant parts in the documentation i.e. to what issues the review should be concentrating. In the detailed review STUK did have difficulties to assess the post-closure safety regarding some of the elements in the safety case because of the extent and structure of the safety case and also explicit safety argumentation that was partly missing in the safety case. This was seen understandable due the uniqueness and first-of-a-kind post-closure safety case concerning disposal of spent nuclear fuel in Finland. Furthermore, because of missing of some key parts in the documentation those were noticed in the completeness review the detailed review of these parts of the documentation was delayed. 70

71 The explicit safety argumentation has been intrinsic value and has long tradition and lot of experience in the nuclear reactor safety assessments. Similar explicit safety argumentation and precise minimum content and structure of the safety case that have been agreed before hand between the regulator and the license applicant should be introduced to the nuclear waste post-closure safety cases also. That would make post-closure safety case easier to review for the regulator and increase traceability and transparency. STUK decided to reform the regulatory guidance for nuclear facilities in The new guidance was planned to be published well before Posiva was going to submit the CLA. Publication was postponed because of delays in the project and Fukushima accident and thus the new YVL guides were released after the submittal of the CLA. Despite of this, CLA was written based on the new guidance. This was possible because STUK agreed with Posiva that Posiva would use final draft versions of the guidance for the CLA documentation. A few late changes in the guidance caused some challenges for the license applicant to prepare the documentation as well as for STUK for the reviewing it. Thus it can be concluded that the guidance should be ready for use for the license applicant well in advance before the submission of the license application and any major changes in the guidance should not be done during the review but afterwards. As a whole it was considered important that the CLA was reviewed against the latest regulations. It should be also clear that there is a mutual understanding between the regulator and the license applicant concerning the requirements. During the detailed review phase STUK made approximately 30 requests for additional information in areas where further information or clarification was needed. Requests for additional information concerned mainly main documents e.g. PSAR and post-closure safety case. As a result of the detailed review phase STUK accepted main documents e.g. PSAR and post-closure safety case and submitted statement and safety evaluation report [1] to the Government. STUK s main conclusion was that encapsulation plant and disposal facility can be built to be safe. Also there is a sufficient reliability that there will be no detrimental radiation effects to the public or environment neither during the operational period nor after decommissioning and closure of the facility. In the statement to the government STUK raised areas that need further development before specific construction step or before submittal of operating license application. These areas are related for example to process for selecting suitable disposal tunnel location, further R&D and assessment of engineered barrier system performance and development in post-closure scenario analysis and presentation of post-closure safety case. These areas have been further specified in PSAR and post-closure safety case decisions [2, 3] that STUK has send directly to Posiva. Raising these areas in the statement and decisions STUK created more steps to already step-wise licensing process. This is justified due the uniqueness and first-of-a-kind nature of this kind of nuclear waste disposal concept. 5. Conclusions A well planned preparation for the review during pre-licensing phase is a key factor to a successful review. Plan for the review should be prepared and resources assured before the review. The review should be guided as detailed as possible. Completeness review before the detailed review phase outlines the scope of the safety case and identifies the shortcomings in it. The review process revealed that more thorough discussions between the regulator and the license applicant before the review would have been needed. Especially more thorough discussions concerning the structure of the documentation and the level of details would have 71

72 simplified and facilitated the review process. Regulations that are up to date are of importance for this kind of new nuclear waste disposal concept. Also discussions concerning the mutual understanding on the regulatory guidance are essential for avoidance of any misunderstandings between the regulator and the license applicant. However, despite the appeared challenges the review could be concluded because of the active communication with the implementer during the review phase. Lessons learned during the review process have been analyzed and development of review practices and safety guidance have been started. REFERENCES [1] STUK, STUK s statement and safety assessment on the construction of the Olkiluoto encapsulation plan and disposal facility for spent nuclear fuel, STUK-B 196, Helsinki [2] STUK, STUK s decision on the PSAR of the encapsulation plant and disposal facility for spent nuclear fuel (in Finnish), , Helsinki. [3] STUK, STUK s review on the construction license stage post closure safety case of the spent nuclear fuel disposal in Olkiluoto, STUK-B 197, Helsinki

73 03d 16 / ID 155. Disposal of High Level Waste GENERIC UNDERGROUND RESEARCH FACILITY IN THE MIDDLE STAGE OF THE SITE SELECTION PROCESS: BUKOV URF, CZECH REPUBLIC L. Vondrovic, I. Pospíšková, J. Augusta, J. Slovák, A. Vokál Radioactive Waste Repository Authority contact of main author: vondrovic@surao.cz Abstract. The Czech Republic s radioactive waste disposal concept assumes the construction of a deep geological repository in crystalline host rocks (granitic and metamorphic) at a depth of 500m below the earth s surface. The current stage of the site selection and evaluation process requires that the characteristics of the geosphere be determined at a depth envisaged for the future repository. This abstract addresses the current state of construction and preparation of the R&D programme for the new Bukov generic underground research facility. This facility will, over the next 10 years or so, provide invaluable support for the current siting process and the safety evaluation of the disposal concept in the Czech Republic by providing the depth calibration parameters required to supplement the data acquired from surface exploration. The Bukov URF (located within the Rožná uranium mine complex) is located at a depth of 600m in metamorphic rocks in the proximity of a potential site for the construction of the future DGR. The laboratory itself is currently approaching the end of the construction phase which commenced in The intensive characterization programme which was conducted during the construction phase focused on the characterization of the site from the geological, geomechanical and hydrogeological aspects. The data set acquired from this initial scientific programme will serve as input data for the construction of synthetic geosphere models which, in turn, will serve for determining the precise location of the various experiments and the development of specialised site description methodologies; moreover, it will provide essential information concerning the design of the future repository. One of the most important parts of the characterization programme will consist of the long-term monitoring of the geological processes that take place at repository depth within the Bohemian Massif. The future experimental programme will focus on the following principal research areas: the geosphere and materials and techniques. It is anticipated that the geosphere part of the programme will provide a description of the characteristics of rock mass behaviour in terms of migration properties, fracture connectivity and the long-term stability of the rock mass as well as the stability of the underground construction itself. The materials part will focus on the long-term stability of the various materials employed and their degradation rates. Finally, the technology part will provide valuable information concerning the preparation of individual components of the emplacement system and the support infrastructure. Key Words: URF, disposal, experimental programme 1. Introduction The SURAO generic research programme is focused on the detailed testing of the crystalline rock concept. Generic laboratories serve as training centres for staff members, experimentation involving mock-up experiments and the development of methodologies for the study of rock conditions in underground environments. One of the most important aspects of generic research consists of the testing of the validity of data collected from the earth s surface and the approximation of such data to depths at which the construction of the repository is envisaged. SURAO has close connections with three underground research centres in the Czech Republic: the Josef Gallery, the Bedřichov Water Supply Tunnel and the Bukov Underground Research Facility (Bukov URF). The Bukov underground generic laboratory is located in the eastern part of the Czech Republic near the Kraví hora candidate repository site and adjacent to the Rožná uranium mine at a depth of 600m below the earth s surface. From the geological point of view the facility is located in the north eastern part of the Moldanubian Zone of the Variscan orogen and is composed of migmatitised paragneisses 73

74 with amphibolite layers. The felsic granulites display the same deformational history as that of the nearby Kraví hora candidate locality. 2. Construction FIG. 1. Scheme of URF Bukov Construction commenced in 2013 with the blasting of the main access tunnel. Following an intensive drilling campaign (the total length of the boreholes amounted to 500m) two suitable rock blocks were defined for testing purposes: a consolidated block made up of high quality rock intended for diffusion and demonstration experiments and slightly fractured rock for the performance of migration, hydraulic and material tests. The laboratory itself consists of a 300m-long connecting cross gallery with a profile of 9.2m2 leading from the access shaft and the underground facility itself consisting of a 90m-long large-profile chamber and a gallery niche system with a total length of 40m (see Fig 1). The second test chamber section consists of a 20m-long niche in the front part of the access tunnel. Rock bolts will be used to provide support for the underground sections supplemented with yieldable TH arches in areas exhibiting more complicated geological conditions. The intensive drilling campaign consisted of the drilling of a series of exploration boreholes of a total length of 500m into the walls of the access tunnel for geophysical and hydrogeological monitoring purposes. The conventional blasting method was used for the excavation of the access tunnel, whereas the smooth blasting method was applied with respect to the laboratory niches. 3. Characterization phase The scientific programme conducted during the construction of the facility concentrated on the characterization of the site from the geological, geomechanical and hydrogeological points of view. The results will serve as input material for the construction of synthetic geosphere 74

75 models which will, in turn, serve for the precise positioning of the various experiments included in the research programme. The characterization programme will include the following research areas: Complex geological characterization The application of a range of geological methods will be aimed at obtaining a multidisciplinary description of the host rock in order to assist in determining the optimum location for the performance of the experimental programme. Geological characterization will comprise geological and structural mapping and the deciphering of the temporal, spatial and thermal evolution of the ductile and brittle pattern. Subsequent more detailed characterization will concentrate on more specialized study fields e.g. the radiometric dating of the fault system, the evolution of micro-fractures within the rock, etc. Geotechnical characterization The geotechnical programme will be made up of three specific areas: (i) stress monitoring, (ii) geotechnical laboratory testing and (iii) seismic monitoring. The stress measurements will allow for the prediction of the stability of the rock mass as well as for the determination of stress changes during the excavation process. Geotechnical testing will comprise a range of methods that will serve for initial rock mass characterization purposes and for the provision of input data for further geotechnical modelling. Seismic monitoring will be concerned with the potential reactivation of the fault system during blasting and the identification of any induced seismic activity that might occur as a result of local mining operations. Transport properties of the rocks The determination of the transport properties of the surrounding rock will serve for the laboratory testing of radionuclide sorption and migration from a depth at which the construction of the repository is envisaged. Hydrogeological properties of the rock mass An understanding of the behaviour of water within the repository system is crucial in terms of safety case considerations. Hydrogeological studies therefore include the monitoring of water influx and the evolution of the chemical and physical properties of water collected from the surrounding rock. Borehole hydrogeological tests, tracer tests and water pressure tests will be conducted during the experimental phase. Synthetic geosphere models The application of the methods described above will result in the construction of the following synthetic geosphere models: 3D structural-geological model 3D hydrogeological model 3D geotechnical model 4. Experimental programme The underground research facility research and experimental (R&E) programme will be conducted in very similar conditions to those expected at the location of the future deep geological repository. The Bukov URF will serve as a test site for assessing the behaviour of the rocks at the candidate sites at a depth matching the expected depth of the deep geological repository until the final site is selected and the confirmation underground laboratory is built at that site. 75

76 The experimental programme consists of 7 basic areas: R&E Programme 1: Pilot characterization of the rocks in order to test the methodology for setting up 3D Geo / GT / HG models of the site R&E Programme 2: Testing of long-term monitoring methods for processes occurring at repository depth R&E Programme 3: Testing of groundwater flow / radionuclide transport models of the fracture environment of the DGR R&E Programme 4: Testing of the effect of the rock at repository depth on the properties of the engineered barriers R&E Programme 5: Testing of the development of excavation disturbed/damaged zones in crystalline complex rocks at repository depth R&E Programme 6: Investigation of the effect of the rock massif on the underground structures of the DGR R&E Programme 7: Demonstration experiments 5. Conclusion The construction of the Bukov Underground Research Facility is fundamental in terms of the characterization of rock masses in which it is intended that the future Czech radioactive waste repository will be constructed. The facility is ideally located for this purpose, i.e. it is 600m beneath the earth s surface and situated in a crystalline rock environment. The research to be conducted at the facility will make a significant contribution towards forming a more detailed understanding of the processes that will take place within the repository over its lifetime. 76

77 03d 17 / ID 161. Disposal of High Level Waste CIGEO PROJECT: FROM BASIC DESIGN TO DETAILED DESIGN PPURSUANT TO REVERSIBILITY F. Launeau, G. Ouzounian Andra, French National Radioactive Waste Management Agency, Parc de la Croix Blanche, Châtenay-Malabry, France contact of main author: frederic.launeau@andra.fr Abstract. The Cigeo project has been in the works for 25 years. Numerous studies have been conducted, with further specific research thanks to direct access to the Callovo-Oxfordian clay formation from the underground laboratory of Bure-Saudron. These studies and research initially aimed to demonstrate the feasibility of the repository. They also helped gain a high level of understanding of phenomena to support design studies and demonstrate safety. Transition to the industrial phase began with the development of a plan for delivering waste to the facility for disposal. The plan introduced sequencing for the various types of waste to be disposed of, and was optimised to determine the size of inspection, transfer and handling facilities. In describing the life of the repository and therefore the vision for its operation, it has become obvious that our generation should not impose choices on future generations. We must provide them with reference technical solutions, with the financial resources to implement them. It is also our duty to begin the construction and initial operating phases. However, because the facility will operate over more than 5 generations, we must leave a degree of flexibility so that they may reassess the options that we define and adopt their own solutions, as necessary. They will also benefit from operational experience gathered as facility operations develop. This is the context in which the preliminary design phase was finalised in preparation for the detailed design phase, with the aim of gradual commissioning during the latter part of the next decade. 1. Introduction Since the Act of December 1991 concerning research into the management of radioactive waste, Andra has been conducting the programme for geological disposal in compliance with the objectives set forth. The initial 15-year phase was mainly dedicated to research, including research into alternatives to geological disposal. Following the various bids for the creation of an underground laboratory, in 1998 the French Government selected the Bure-Saudron facility in the Meuse and Haute-Marne departments of north-eastern France. In 2005, Andra compiled the results and analysed them in the Dossier 2005 Argile report. The main finding of Dossier 2005 was that geological disposal is feasible in the clay formation studied (Callovo- Oxfordian clay) and that its safety could be proven. Based on the various results, French Parliament passed the Planning Act in 2006, establishing geological disposal as the reference solution for managing high-level waste (HLW) and intermediate-level long-lived waste (ILW- LL). The facilities should be planned in a formation previously studied using an underground laboratory, which indicates the Callovo-Oxfordian near Bure-Saudron. More detailed investigations therefore focused on this region and in 2009, Andra proposed the location for underground facilities. Upon completion of a series of assessments and opinions, the French Government validated the location for the underground repository in March This began the industrialisation process for the Cigeo project, followed by a public debate in 2013, which became useful for later deliberations. When the preliminary design phase was completed and before beginning the detailed design phase, the life and operation of the disposal facility were reviewed using updated information to bring a new perspective to the industrial project. Due to changes to regulatory requirements in France, Cigeo s detailed design must be used for the repository construction license application. The construction license application will therefore 77

78 be submitted progressively between late 2015 and mid 2018 in agreement with safety authorities. 9. Development of Cigeo Project Based on this initial research, Andra proposed an initial project in 2001, which was followed by a detailed safety assessment. This was submitted for international review and created the basis for Dossier The demonstration provided was supported by an understanding of the phenomena affecting the behaviour of the repository gained from a sustained research effort. The repository was no longer viewed as a single object placed in the geological environment, but rather as a group of structures and components developing over time and subject to relatively complex physical-chemical and sometimes combined phenomena. The approach, now called Phenomenological Analysis of Repository Situations, has demonstrated an unparalleled ability to describe repository operation. Based on this analysis, new developments and improvements to the characteristics of the structures and components were made. An overall architecture was developed as a working basis to begin the initial industrial development phases. Once the location of the future repository was known, more detailed drawings were produced, thus validating the overall architecture comprising: Surface nuclear facilities used for receiving, inspecting and conditioning waste, then transferring packages underground via a funicular; An approximately 4.2 km long ramp to transfer surface waste packages underground; A surface mining facility, including access shafts to underground facilities; An underground facility with a disposal area for ILW-LL, and a disposal area for high-level vitrified waste. In 2010, this overview of the main options was confirmed. Based on these main options, the design phase began, particularly with the preparation of the preliminary design. A first draft was submitted for public debate in It was used as the basis for later discussions with local and regional representatives concerning the location of surface facilities. After public debate, the location was decided. Several possible zones were identified directly below surface facilities for mining activities. Local representatives preferred wooded areas in order to avoid encroaching on farmland. For nuclear facilities, the planned sector is located directly next to the underground laboratory, straddling the border between the Meuse and Haute-Marne departments. 3. Launch of the design phase The technical feasibility of the geological repository relied on simple, robust technical concepts. Studies and research conducted since have explored avenues for optimisation and provided more specific details for the basic options in order to develop a preliminary design for a disposal facility. The Cigeo geological repository must be able to hold a wide variety of waste packages, particularly those generated from decades of research and development of industrial processes. Packages will include cemented intermediate-level waste, bituminised waste, and packages in various forms with different characteristics. To simplify operations, the various packages were divided into types for which disposal packages had to be developed. System standardization has been implemented via use of disposal containers. 78

79 The inventory of waste to be disposed of in Cigeo includes 10,000 m 3 of vitrified high-level waste and 70,000 m 3 of ILW-LL. The repository is therefore designed to be large enough to hold this inventory, and operating facilities must be capable of handling the waste and emplacing it in the repository. The repository architecture groups together the disposal cells for different waste categories within specific repository zones. ILW-LL and HLW repository zones will therefore be physically separated from one another. This will ensure phenomenological independence between each zone over the long term. Disposal zones will be built gradually in successive phases, as new packages are received. They will therefore be designed in modules. During the operation of the repository, surface facilities will manage waste packages before they are transferred to underground disposal facilities. They will also support underground operations. These facilities are designed to be decommissioned when the closure decision is made. 4. Cigeo lifecycle phase and governance The main, successive phases of the Cigeo project are as follows: 1. facility "design", including the technical specification of the facility structures, buildings and procedures. This phase ends with the completion of detailed design and the construction license application; Subject to authorisation by decree (construction license): 2. initial construction of Cigeo when the first part of the facility is built. This includes surface buildings associated with operation of the surface nuclear facility, surface-tobottom connections and underground structures to receive the first waste packages; 3. following issue of the operating license for Cigeo, operation by successive phases over around one hundred years with package acceptance and disposal carried out in parallel to underground facility extension work, in order to continue acceptance of packages in the inventory. Partial closure work (moving to Stages 3 and 4 on the International Retrievability Scale) is also carried out in addition to construction, adaptation and regeneration work on surface buildings; 4. the pilot industrial phase planned for the launch of Cigeo operation before the switch to normal operation. This pilot industrial phase will include tests designed to demonstrate the ability to remove waste packages disposed of in Cigeo under real conditions; after operation has finished, the decommissioning and final closure of Cigeo, which can only be authorised by the passing of an Act of Parliament. Cigeo then enters its monitoring phase. Construction and operation will be gradually developed in line with the forecasts for waste package delivery. 5. Gradual development Pursuing the process of creating a deep geological disposal facility is an ethical obligation for our generation as important as ensuring that coming generations are able to reconsider any decisions taken. In both instances, it is about not committing these generations to the choices we make or fail to make. It is our generation and the previous one which built nuclear power plants and enjoyed the benefits in terms of development and lifestyle. We must therefore bear the investment cost for managing the waste produced. The technology and financial resources 79

80 required to carry out the first stages of Cigeo development are now available. Nuclear power plants are still in operation and will continue to support the funding of future investment phases in the medium term. By gradually implementing Cigeo, it is possible both to prepare for disposal of the HLW that produces the most heat and to avoid any time gaps in waste management throughout the Cigeo operation period. It should be noted that the very first vitrified waste packages produced in the 1970s will be sent for initial highly instrumented disposal, in order to prepare for the highly exothermic vitrified waste packages from Reversibility and tools The ethical concern for reversibility comes from the time scale required for managing the most harmful radioactive waste. Particularly given the planned duration of approximately 120 years for the geological disposal facility operation, it is our generation s responsibility to design and provide future generations with a safe facility that they will be able to modify or improve in accordance with their own objectives and requirements, or even replace by other management facilities if other choices become available, particularly due to technical advances. The reversibility of disposal is considered to be the ability to leave the next generation choices concerning the long-term management of radioactive waste, including the choice of reconsidering the decisions made by the previous generation. In practice, reversibility is based on governance tools and technical project management tools Governance tools: continuous improvement of understanding of radioactive waste management, transparency and passing down of information and knowledge, the involvement of society and checks by the government and assessment bodies. Project management tools: incremental development and gradual approach to the construction of Cigeo facilities, flexible operation, adaptability of facilities and retrievability of packages. These tools support decision-making for radioactive waste management. In particular, they ensure that the various choices available are preserved or unlocked over time. With this new understanding of operation, retrievability is simply a technical possibility given to the following generations so that they can implement their own options. To this end, our responsibility is to provide facilities that are designed from the offset to be able to reconsider our choices at a later time if required. As well as passing down high-quality options, we are offering the necessary funds for their implementation. However, future generations will have to bear the cost of any changes in direction. 7. Conclusion The vision of the Cigeo project had long remained fairly static. It had been about creating an overview with the aim of carrying out phenomenological studies and many safety analyses in the long term. These steps have been completed, in particular between the promulgation of French Acts of 1991 and As the industrial phase approaches, the vision is becoming increasingly dynamic, incorporating designers in the disposal lifecycle. Disposal operation will be carried out very gradually, in the frame of pilot industrial phase starting by trials in the mid-2020s and a completion of commissioning in the mid-2030s. 80

81 03d 18 / ID 171. Disposal of High Level Waste IMPACT OF STORAGE PERIOD ON SAFE GEOLOGICAL DISPOSAL OF SPENT FUEL B.B. Acar 1, H.O. Zabunoğlu 2 1 Turkish Atomic Energy Authority, Ankara, Turkey 2 Department of Nuclear Engineering, Hacettepe University, Ankara,Turkey contact of main author: banubulut.acar@taek.gov.tr Abstract. Geological disposal is the widely accepted method for safe final disposal of spent fuel (SF) and high level waste (HLW). Currently, there are no active deep geological repositories. However, various geological disposal projects are under way in many countries. In geological disposal, canisters containing SF/HLW are simply placed into boreholes in a geological formation deep underground, specifically selected for final disposal of nuclear wastes. The main factor affecting the geological repository design is the amount of waste that can be safely emplaced per unit area of the repository (waste disposal density) and it strongly depends on the characteristics (amount, isotopic composition, heat generation rate etc.) of the waste. The isotopic composition and heat generation rate of SF discharged from reactor change during storage. This study aims to assess the effect of interim storage period on disposal density of SF in a geological repository. In the first part of the study, utilizing the code Monteburns, relevant compositions and decay heats of SFs discharged from a reference PWR (A 1000-MWe PWR loaded with 3.3 w/o enriched UO 2 fuel, with a discharge burn up of MWd/tU and with an irradiation time of 1000 days) are obtained for selected cooling times. Then, using the code ANSYS, thermal analyses are performed for a reference repository concept and disposal areas needed for SFs with different ages are determined by ensuring that thermal criteria limiting the canister surface temperature is satisfied. Results of the analysis are used to assess the effect of storage period of SF on disposal layout and to derive the correlation between storage period and safe disposal capacity of geological repository. Key Words: Spent fuel, geological disposal, storage, disposal density. 1. Introduction Heat dissipation from a radioactive waste is one of the most important factors in geological repository design and it depends on the waste type and composition. Waste composition is a function of enrichment and burnup of the fuel, reactor power and cooling time of waste. Disposal density calculations have two major parts: (1) determination of compositions and decay heat profiles of wastes and (2) determination of disposal area through thermal analysis. 2. Determination of Characteristics of Spent Fuels In this part of the study, isotopic compositions and decay heat profiles of SFs with different storage periods are evaluated for a reference PWR by using Monteburns code. Monteburns is a fully automated tool that links the Monte Carlo transport code MCNP with the radioactive decay and burnup code Origen2. Monteburns produces a large number of criticality and burnup results based on various material feed/removal specifications, power(s), and time intervals. In this study, inputs are prepared with reference reactor technical data and by unit cell approximation. 2.1.Reference Reactor A 1000-MWe PWR loaded with 3.3 w/o enriched UO 2 fuel, with discharge burnup of MWd/tU and with an irradiation time of 1000 days is taken as the reference. SF discharged in 81

82 the reference case consists of about 95.5 w/o U, 1 w/o Pu, 3.5 w/o fission products and other actinides. The U in SF contains around 0.85 w/o U-235. About 70 w/o of Pu in SF is composed of fissile isotopes (~59 w/o Pu-239 and ~11 w/o Pu-241). 2.2.Decay Heat Profiles Decay heat profile of SF is obtained from Monteburns output for 10 6 years decay period. This decay heat profile is shown in Figure 1 and is used as source term in thermal analysis. FIG. 1.Decay heat of SF with MWd/tHM burnup In order to obtain heat generation rate equations for SFs with different storage periods (40, 50, 60, 80 and 100 years) which are to be used as heat source terms in thermal analyses, time dependent decay heat curve is fitted to sum of four exponential terms «Put s formula»[1]: Where Q is decay heat in W/tHM; t is time in year elapsed since the production of the SF. Values of the coefficients to be used in Put s formula are given in Table I. TABLE I: VALUES OF THE COEFFICIENTS IN PUT'S FORMULA A 1 A 2 A 3 A 4 b 1 b 2 b 3 b E-5 3. Disposal Density Calculations Once SFs disposed in the repository, temperatures of the repository components increase due to the heat generation. Temperature affects many processes occurring in the repository, thus, during the repository design, it is necessary to determine an appropriate density of emplacement of heat-generating wastes and investigate the resultant time-dependent temperature distributions. 3.1.Reference Repository Concept Q t Ai e The KBS-3 concept developed by Swedish Nuclear Fuel and Waste Company is taken as the reference repository. In the reference disposal concept, SF is placed into copper canisters with a cast iron insert. The canisters are surrounded by bentonite buffer and placed vertically into holes in parallel tunnels at a depth of 500 m in granite rock. The depth of hole for SF canister is 7.55 and the diameter of hole is 1.75 meters. Tunnel diameter is 5.5 meters. The distance between the tunnels is 40 meters [2]. Four SF assemblies would be packaged within a copper canister. Each SF assembly has a square cross-section m by m and 4.1 m long. i bit 82

83 Disposal canister is 4.5 m long and 0.9 m in diameter [2]. Figure 2 shows reference repository concept. 3.2.Thermal Analysis FIG. 2.Reference repository concept [2] and SF disposal canister Once disposal canisters are disposed in the repository, a transient heat diffusion phenomenon gives rise because of the heat generated in disposal canisters. Heat transfer in the repository is mainly by conduction. ANSYS finite element code is used to develop a 3-D thermal model of the repository. It is assumed that the repository contains infinite number of tunnels filled with infinite number of canisters with the same thermal output. Due to the geometrical and loading symmetry of the repository, thermal model is simplified to one quarter of a deposition hole with three symmetry surfaces. Vertical symmetry planes passing through the center of the holes, half distance between the adjacent holes and half distance between the adjacent tunnels constitute the lateral boundaries of the model. Figure 3 shows ANSYS model of repository. FIG. 3.ANSYS model of repository Constant temperature boundary conditions are applied at the top and bottom boundaries of the model. All symmetric boundaries are assumed to be adiabatic. The heat-source term is applied as volumetric heat generation in the waste region. Thermal analyses are performed for various spacing values and the minimum distance between boreholes is determined with reference to the thermal constraint. The thermal constraint is that the temperature at the canister surface must not exceed 100 ºC. Bentonite will remain chemically intact for more than one million years as long as the temperature does not exceed 100 ºC [3]. In this study, the temperature limit is reduced to 80 ºC, in order to include a margin of 10 ºC to cover for natural deviations in environmental parameters and another 10 ºC to cover the risk of occurrence of an air gap 83