Training courses on integrated safety assessment modelling for waste repositories

International Nuclear Information System (INIS)

Mallants, D.

2007-01-01

Near-surface or deep repositories of radioactive waste are being developed and evaluated all over the world. Also, existing repositories for low- and intermediate-level waste often need to be re-evaluated to extend their license or to obtain permission for final closure. The evaluation encompasses both a technical feasibility as well as a safety analysis. The long term safety is usually demonstrated by means of performance or safety assessment. For this purpose computer models are used that calculate the migration of radionuclides from the conditioned radioactive waste, through engineered barriers to the environment (groundwater, surface water, and biosphere). Integrated safety assessment modelling addresses all relevant radionuclide pathways from source to receptor (man), using in combination various computer codes in which the most relevant physical, chemical, mechanical, or even microbiological processes are mathematically described. SCK-CEN organizes training courses in Integrated safety assessment modelling that are intended for individuals who have either a controlling or supervising role within the national radwaste agencies or regulating authorities, or for technical experts that carry out the actual post-closure safety assessment for an existing or new repository. Courses are organised by the Department of Waste and Disposal

Climate Considerations in Long-Term Safety Assessments for Nuclear Waste Repositories

Energy Technology Data Exchange (ETDEWEB)

Naeslund, Jens-Ove; Brandefelt, Jenny; Claesson Liljedahl, Lillemor [Svensk Kaernbraenslehantering AB, Stockholm (Sweden)], E-mail: jens-ove.naslund@skb.se

2013-05-15

For a deep geological repository for spent nuclear fuel planned in Sweden, the safety assessment covers up to 1 million years. Climate scenarios range from high-end global warming for the coming 100 000 years, through deep permafrost, to large ice sheets during glacial conditions. In contrast, in an existing repository for short-lived waste the activity decays to low levels within a few tens of thousands of years. The shorter assessment period, 100 000 years, requires more focus on climate development over the coming tens of thousands of years, including the earliest possibility for permafrost growth and freezing of the engineered system. The handling of climate and climate change in safety assessments must be tailor-made for each repository concept and waste type. However, due to the uncertain future climate development on these vast time scales, all safety assessments for nuclear waste repositories require a range of possible climate scenarios.

The deep geologic repository technology programme: toward a geoscience basis for understanding repository safety

International Nuclear Information System (INIS)

Jensen, M.R.

2007-01-01

Within the Deep Geologic Repository Technology Programme (DGRTP) several Geoscience activities are focused on advancing the understanding of groundwater flow system evolution and geochemical stability in a Canadian Shield setting as affected by long-term climate change. A key aspect is developing confidence in predictions of groundwater flow patterns and residence times as they relate to the safety of a deep geologic repository for used nuclear fuel waste. This is being achieved through a coordinated multi-disciplinary approach intent on: i) demonstrating coincidence between independent geo-scientific data; ii) improving the traceability of geo-scientific data and its interpretation within a conceptual descriptive model(s); iii) improving upon methods to assess and demonstrate robustness in flow domain prediction(s) given inherent flow domain uncertainties (i.e. spatial chemical/physical property distributions, boundary conditions) in time and space; and iv) improving awareness amongst geo-scientists as to the utility of various geo-scientific data in supporting a safety case for a deep geologic repository. This multi-disciplinary DGRTP approach is yielding an improved understanding of groundwater flow system evolution and stability in Canadian Shield settings that is further contributing to the geo-scientific basis for understanding and communicating aspects of DGR safety. (author)

Climate considerations in long-term safety assessments for nuclear waste repositories.

Science.gov (United States)

Näslund, Jens-Ove; Brandefelt, Jenny; Liljedahl, Lillemor Claesson

2013-05-01

For a deep geological repository for spent nuclear fuel planned in Sweden, the safety assessment covers up to 1 million years. Climate scenarios range from high-end global warming for the coming 100 000 years, through deep permafrost, to large ice sheets during glacial conditions. In contrast, in an existing repository for short-lived waste the activity decays to low levels within a few tens of thousands of years. The shorter assessment period, 100 000 years, requires more focus on climate development over the coming tens of thousands of years, including the earliest possibility for permafrost growth and freezing of the engineered system. The handling of climate and climate change in safety assessments must be tailor-made for each repository concept and waste type. However, due to the uncertain future climate development on these vast time scales, all safety assessments for nuclear waste repositories require a range of possible climate scenarios.

Repository Safety Strategy: Strategy for Protecting Public Health and Safety after Closure of a Yucca Mountain Repository, Rev. 1

Energy Technology Data Exchange (ETDEWEB)

DOE

1998-01-01

The updated Strategy to Protect Public Health and Safety explains the roles that the natural and engineered systems are expected to play in achieving the objectives of a potential repository system at Yucca Mountain. These objectives are to contain the radionuclides within the waste packages for thousands of years, and to ensure that annual doses to a person living near the site will be acceptably low. This strategy maintains the key assumption of the Site Characterization Plan (DOE 1988) strategy that the potential repository level (horizon) will remain unsaturated. Thus, the strategy continues to rely on the natural attributes of the unsaturated zone for primary protection by providing a setting where waste packages assisted by other engineered barriers are expected to contain wastes for thousands of years. As in the Site Characterization Plan (DOE 1988) strategy, the natural system from the walls of the underground openings (drifts) to the human environment is expected to provide additional defense by reducing the concentrations of any radionuclides released from the waste packages. The updated Strategy to Protect Public Health and Safety is the framework for the integration of site information, repository design and assessment of postclosure performance to develop a safety case for the viability assessment and a subsequent license application. Current site information and a reference design are used to develop a quantitative assessment of performance to be compared with a performance measure. Four key attributes of an unsaturated repository system that are critical to meeting the objectives: (1) Limited water contacting the waste packages; (2) Long waste package lifetime; (3) Slow rate of release of radionuclides from the waste form; and (4) Concentration reduction during transport through engineered and natural barriers.

A strategy to establish Food Safety Model Repositories.

Science.gov (United States)

Plaza-Rodríguez, C; Thoens, C; Falenski, A; Weiser, A A; Appel, B; Kaesbohrer, A; Filter, M

2015-07-02

Transferring the knowledge of predictive microbiology into real world food manufacturing applications is still a major challenge for the whole food safety modelling community. To facilitate this process, a strategy for creating open, community driven and web-based predictive microbial model repositories is proposed. These collaborative model resources could significantly improve the transfer of knowledge from research into commercial and governmental applications and also increase efficiency, transparency and usability of predictive models. To demonstrate the feasibility, predictive models of Salmonella in beef previously published in the scientific literature were re-implemented using an open source software tool called PMM-Lab. The models were made publicly available in a Food Safety Model Repository within the OpenML for Predictive Modelling in Food community project. Three different approaches were used to create new models in the model repositories: (1) all information relevant for model re-implementation is available in a scientific publication, (2) model parameters can be imported from tabular parameter collections and (3) models have to be generated from experimental data or primary model parameters. All three approaches were demonstrated in the paper. The sample Food Safety Model Repository is available via: http://sourceforge.net/projects/microbialmodelingexchange/files/models and the PMM-Lab software can be downloaded from http://sourceforge.net/projects/pmmlab/. This work also illustrates that a standardized information exchange format for predictive microbial models, as the key component of this strategy, could be established by adoption of resources from the Systems Biology domain. Copyright © 2015. Published by Elsevier B.V.

The study on safety facility criteria for radioactive waste repository

International Nuclear Information System (INIS)

Lee, S. H.; Choi, M. H.; Han, S. H. and others

1992-12-01

The radioactive waste repository are necessary to install the engineered safety systems to secure the safety for operation of the repository in the event of fire and earthquake. Since the development of safety facility criteria requires a thorough understanding about the characteristics of the engineered safety systems, we should investigate by means of literature survey and visit SKB. In particular, definition, composition of the systems, functional requirement of the systems, engineered safety systems of foreign countries, system design, operation and maintenance requirement should be investigated : fire protection system, ventilation system, drainage system, I and C system, electric system, radiation monitoring system. This proposed criteria consist of purpose, scope of application, ventilation system, fire protection system, drainage system, electric system and this proposed criteria can be applied as a basic reference for the final criteria

Postclosure safety assessment of a used fuel repository in sedimentary rock

International Nuclear Information System (INIS)

Gobien, M.; Garisto, F.; Hunt, N.; Kremer, E.

2014-01-01

The Nuclear Waste Management Organization (NWMO) is responsible for the implementation of Adaptive Phased Management (APM), the federally-approved plan for safe long-term management of Canada's used nuclear fuel. Under the APM plan, used nuclear fuel will ultimately be placed within a deep geological repository in a suitable rock formation. This paper summarizes an illustrative case study of the current multi-barrier design and postclosure safety of a deep geological repository in a hypothetical sedimentary Michigan Basin setting. The purpose of this postclosure safety assessment is to determine potential effects of the repository on the health and safety of persons and the environment. Results are compared against acceptance criteria established for the protection of persons and the environment from potential radiological and non-radiological hazards. (author)

Postclosure safety assessment of a used fuel repository in sedimentary rock

Energy Technology Data Exchange (ETDEWEB)

Gobien, M.; Garisto, F.; Hunt, N.; Kremer, E. [Nuclear Waste Management Organization, Toronto, ON (Canada)

2014-07-01

The Nuclear Waste Management Organization (NWMO) is responsible for the implementation of Adaptive Phased Management (APM), the federally-approved plan for safe long-term management of Canada's used nuclear fuel. Under the APM plan, used nuclear fuel will ultimately be placed within a deep geological repository in a suitable rock formation. This paper summarizes an illustrative case study of the current multi-barrier design and postclosure safety of a deep geological repository in a hypothetical sedimentary Michigan Basin setting. The purpose of this postclosure safety assessment is to determine potential effects of the repository on the health and safety of persons and the environment. Results are compared against acceptance criteria established for the protection of persons and the environment from potential radiological and non-radiological hazards. (author)

Making the Postclosure Safety Case for the Proposed Yucca Mountain Repository

Energy Technology Data Exchange (ETDEWEB)

P. Swift; A.V. Luik

2006-08-28

The International Atomic Energy Agency (IAEA), in its advisory standard for geological repositories promulgated jointly with the Nuclear Energy Agency (NEA) of the Organization for Economic Co-operation and Development, explicitly distinguishes between the concepts of a safety case and a safety assessment. As defined in the advisory standard, the safety case is a broader set of arguments that provide confidence and substantiate the formal analyses of system safety made through the process of safety assessment. Although the IAEAYs definitions include both preclosure (i.e., operational) safety and post-closure performance in the overall safety assessment and safety case, the emphasis in here is on long-term performance after waste has been emplaced and the repository has been closed. This distinction between pre- and postclosure aspects of the repository is consistent with the U.S. regulatory framework defined by the U.S. Environmental Protection Agency (Chapter 40 of the Code of Federal Regulations, Part 197, or 40 CFR 197) [2] and implemented by the U.S. Nuclear Regulatory Commission (Chapter 10 of the Code of Federal Regulations, Part 63, or 10 CFR 63) [3]. The separation of the pre- and postclosure safety cases is also consistent with the way in which the U.S. Department of Energy has assigned responsibilities for developing the safety case. Bechtel SAIC Company is the Management and Operating contractor responsible for the design and operation of the Yucca Mountain facility and is therefore responsible for the preparation of the preclosure aspects of the safety case. Sandia National Laboratories has lead responsibility for scientific work evaluating post-closure performance, and therefore is responsible for developing the post-closure aspects of the safety case. In the context of the IAEA definitions, both preclosure and postclosure safety, including safety assessment and the safety case, will be documented in the license application being prepared for the

Making the Postclosure Safety Case for the Proposed Yucca Mountain Repository

International Nuclear Information System (INIS)

P. Swift; A.V. Luik

2006-01-01

The International Atomic Energy Agency (IAEA), in its advisory standard for geological repositories promulgated jointly with the Nuclear Energy Agency (NEA) of the Organization for Economic Co-operation and Development, explicitly distinguishes between the concepts of a safety case and a safety assessment. As defined in the advisory standard, the safety case is a broader set of arguments that provide confidence and substantiate the formal analyses of system safety made through the process of safety assessment. Although the IAEAYs definitions include both preclosure (i.e., operational) safety and post-closure performance in the overall safety assessment and safety case, the emphasis in here is on long-term performance after waste has been emplaced and the repository has been closed. This distinction between pre- and postclosure aspects of the repository is consistent with the U.S. regulatory framework defined by the U.S. Environmental Protection Agency (Chapter 40 of the Code of Federal Regulations, Part 197, or 40 CFR 197) [2] and implemented by the U.S. Nuclear Regulatory Commission (Chapter 10 of the Code of Federal Regulations, Part 63, or 10 CFR 63) [3]. The separation of the pre- and postclosure safety cases is also consistent with the way in which the U.S. Department of Energy has assigned responsibilities for developing the safety case. Bechtel SAIC Company is the Management and Operating contractor responsible for the design and operation of the Yucca Mountain facility and is therefore responsible for the preparation of the preclosure aspects of the safety case. Sandia National Laboratories has lead responsibility for scientific work evaluating post-closure performance, and therefore is responsible for developing the post-closure aspects of the safety case. In the context of the IAEA definitions, both preclosure and postclosure safety, including safety assessment and the safety case, will be documented in the license application being prepared for the

Safety assessment for a KBS-3H spent nuclear fuel repository at Olkiluoto. Complementary evaluations of safety

International Nuclear Information System (INIS)

Neall, Fiona; Pastina, Barbara; Snellman, Margit; Smith, Paul; Gribi, P.; Johnson, Lawrence

2008-12-01

The KBS-3H design is a variant of the more general KBS-3 method for the geological disposal of spent nuclear fuel in Finland and Sweden. In the KBS-3H design, multiple assemblies containing spent fuel are emplaced horizontally in parallel, approximately 300 m long, slightly inclined deposition drifts. The copper canisters, each with a surrounding layer of bentonite clay, are placed in perforated steel shells prior to deposition in the drifts; the assembly is called the 'supercontainer'. The other KBS-3 variant is the KBS-3V design, in which the copper canisters are emplaced vertically in individual deposition holes surrounded by bentonite clay but without steel supercontainer shells. SKB and Posiva have conducted a Research, Development and Demonstration programme over the period 2002-2007 with the overall aim of establishing whether KBS-3H represents a feasible alternative to KBS-3V. As part of this programme, the long-term safety of a KBS-3H repository has been assessed in the KBS-3H safety studies. In order to focus the safety studies, the Olkiluoto site in the municipality of Eurajoki, which is the proposed site for a spent fuel repository in Finland, was used as a hypothetical site for a KBS-3H repository. The present report is part of a portfolio of reports discussing the long-term safety of the KBS-3H repository. The overall outcome of the KBS-3H safety studies is documented in the summary report, 'Safety assessment for a KBS-3H repository for spent nuclear fuel at Olkiluoto'. The purpose and scope of the KBS-3H complementary evaluations of safety report is provided in Posiva's Safety Case Plan, which is based on Regulatory Guide YVL 8.4 and on international guidelines on complementary lines of argument to long-term safety that are considered an important element of a post-closure safety case for geological repositories. Complementary evaluations of safety require the use of evaluations, evidence and qualitative supporting arguments that lie outside the

Project Guarantee 1985. Final repository for high-level radioactive wastes: Safety report

International Nuclear Information System (INIS)

Anon.

1985-01-01

Disposal of radioactive was involves preventing releases to the biosphere for a long period of time and subsequently limiting the magnitude of releases by means of a series of safety barriers: the waste solidification matrix (borosilicate glass), massive steel canisters in highly compacted bentonite, sealing of void spacer and access routes on repository closure. The geological barriers are formed by the crystalline bed-rock and the overlying sedimentary layers. In order to perform a safety assessment the behaviour of these technical barriers and of the host rock must be understood and this understanding must be translated into quantitative models which allow calculation of repository performance. For the particular case of a Swiss repository, the main criterion is the individual dose limit of 10 mrem/year, which is given in the safety guidelines of the Swiss authorities. The procedure for the safety analysis involves examination of all scenarios which could give rise to radionuclide release from the repository. Qualitative considerations of both the magnitude of their consequences and their likelihood are used in order to identify a restricted number of scenarios for quantitative analysis

Confidence in the long-term safety of deep geological repositories. Its development and communication

International Nuclear Information System (INIS)

1999-01-01

The technical aspects of confidence have been the subject of considerable debate, especially the concept of model validation. The safety case that is provided at a particular stage in the planning, construction, operation or closure of a deep geological repository is a part of a broader decision basis that guides the repository-development process. The basic steps for deriving the safety case at various stages of repository development involve: a safety assessment; and the documentation of the safety assessment, a statement of confidence in the safety indicated by the assessment, and the confirmation of the appropriateness of the safety strategy. The approaches to establish confidence in the evaluation of safety should aim to ensure that the decisions taken within the incremental process of repository development are well-founded. Various aspects of confidence in the evaluation of safety, and their integration within a safety case, are presented in detail in the present report. When communicating confidence in the findings of a safety assessment, clarity in the communication of concepts is always required. Consistent with this requirement, key concepts are specifically defined in the main text of the report. (R.P.)

Safety assessment of geologic repositories for nuclear waste

International Nuclear Information System (INIS)

Bartlett, J.W.; Burkholder, H.C.; Winegardner, W.K.

1977-01-01

Consideration of geologic isolation for final disposition of radioactive wastes has led to the need for evaluation of the safety of the concept. Such evaluations require consideration of factors not encountered in conventional risk analysis: consequences at times and places far removed from the repository site; indirect, complex, and alternative pathways between the waste and the point of potential consequences; a highly limited data base; and limited opportunity for experimental verification of results. R and D programs to provide technical safety evaluations are under way. Three methods are being considered for the probabilistic aspects of the evaluations: fault tree analysis, repository simulation analysis, and system stability analysis. Nuclide transport models, currently in a relatively advanced state of development, are used to evaluate consequences of postulated loss of geologic isolation. This paper outlines the safety assessment methods, unique features of the assessment problem that affect selection of methods and reliability of results, and available results. It also discusses potential directions for future work

Requirements on the provisional safety analyses and technical comparison of safety measures

International Nuclear Information System (INIS)

2010-04-01

The concept of a Geological Underground Repository (SGT) was adopted by the Swiss Federal Council on April 2 nd , 2008. It fixes the goals and the safety technical criteria as well as the procedures for the choice of the site for an underground repository. Those responsible for waste management evaluate possible site regions according to the present status of geological knowledge and based on the safety criteria defined in SGT as well as on technical feasibility. In a first step, they propose geological repository sites for high level (HAA) and for low and intermediate level (SMA) radioactive wastes and justify their choice in a report delivered to the Swiss Federal Office of Energy. The Swiss Federal Council reviews the choices presented and, in the case of positive evaluation, approves them and considers them as an initial orientation. In a second step, based on the possible sites according to step 1, the waste management institution responsible has to reduce the repositories chosen for HAA and SMA by taking into account safety aspects, technical feasibility as well as space planning and socio-economical aspects. In making this choice, safety aspects have the highest priority. The criteria used for the evaluation in the first step have to be defined using provisional quantitative safety analyses. On the basis of the whole appraisal, including space planning and socio-economical aspects, those responsible for waste management propose at least two repository sites for HAA- and SMA-waste. Their selection is then reviewed by the authorities and, in the case of a positive assesment, the selection is taken as an intermediate result. The remaining sites are further studied to examine site choice and the delivery of a request for a design license. If necessary, the requested geological knowledge has to be confirmed by new investigations. Based on the results of the choosing process and a positive evaluation by the safety authorities, the Swiss Federal Council has to

Natural safety indicators and their application to repository safety cases

International Nuclear Information System (INIS)

Miller, B.

2002-01-01

Radiological dose and risk are the standard end-points calculated in all performance assessments. Their calculation requires, however, assumptions to be made for future human behaviour. To complement dose and risk, other safety indicators have been suggested which do not require such assumptions to be made. One proposed set of safety indicators are the concentrations and fluxes of naturally-occurring chemical species in the environment which may be compared with the performance assessment predictions of repository releases. Such comparisons can be valid because both the natural and repository species would occur in the same system and their transport behaviour would be controlled by exactly the same processes at the same rates. Although simple in concept, there is currently no consensus on the most appropriate comparisons to make or on the interpretation of such comparisons. A number of national and international research projects are evaluating this proposed approach, including an IAEA Co-ordinated Research Programme. These projects suggest that that the approach appears to be workable and that it may be a valuable component of a safety case, complementing the dose and risk presentations. Further work is, however, necessary to develop the approach to a level where it may be confidently applied in further performance assessments in a consistent and methodical manner. (author)

Safety assessment for a KBS-3H spent nuclear fuel repository at Olkiluoto. Complementary evaluations of safety

Energy Technology Data Exchange (ETDEWEB)

Neall, Fiona; Pastina, Barbara; Snellman, Margit; Smith, Paul; Gribi, P.; Johnson, Lawrence

2008-12-15

The KBS-3H design is a variant of the more general KBS-3 method for the geological disposal of spent nuclear fuel in Finland and Sweden. In the KBS-3H design, multiple assemblies containing spent fuel are emplaced horizontally in parallel, approximately 300 m long, slightly inclined deposition drifts. The copper canisters, each with a surrounding layer of bentonite clay, are placed in perforated steel shells prior to deposition in the drifts; the assembly is called the 'supercontainer'. The other KBS-3 variant is the KBS-3V design, in which the copper canisters are emplaced vertically in individual deposition holes surrounded by bentonite clay but without steel supercontainer shells. SKB and Posiva have conducted a Research, Development and Demonstration programme over the period 2002-2007 with the overall aim of establishing whether KBS-3H represents a feasible alternative to KBS-3V. As part of this programme, the long-term safety of a KBS-3H repository has been assessed in the KBS-3H safety studies. In order to focus the safety studies, the Olkiluoto site in the municipality of Eurajoki, which is the proposed site for a spent fuel repository in Finland, was used as a hypothetical site for a KBS-3H repository. The present report is part of a portfolio of reports discussing the long-term safety of the KBS-3H repository. The overall outcome of the KBS-3H safety studies is documented in the summary report, 'Safety assessment for a KBS-3H repository for spent nuclear fuel at Olkiluoto'. The purpose and scope of the KBS-3H complementary evaluations of safety report is provided in Posiva's Safety Case Plan, which is based on Regulatory Guide YVL 8.4 and on international guidelines on complementary lines of argument to long-term safety that are considered an important element of a post-closure safety case for geological repositories. Complementary evaluations of safety require the use of evaluations, evidence and qualitative supporting arguments

Preliminary safety analysis of the Baita Bihor radioactive waste repository, Romania

International Nuclear Information System (INIS)

Little, Richard; Bond, Alex; Watson, Sarah; Dragolici, Felicia; Matyasi, Ludovic; Matyasi, Sandor; Naum, Mihaela; Niculae, Ortenzia; Thorne, Mike

2007-01-01

A project funded under the European Commission's Phare Programme 2002 has undertaken an in-depth analysis of the operational and post-closure safety of the Baita Bihor repository. The repository has accepted low- and some intermediate-level radioactive waste from industry, medical establishments and research activities since 1985 and the current estimate is that disposals might continue for around another 20 to 35 years. The analysis of the operational and post-closure safety of the Baita Bihor repository was carried out in two iterations, with the second iteration resulting in reduced uncertainties, largely as a result taking into account new information on the hydrology and hydrogeology of the area, collected as part of the project. Impacts were evaluated for the maximum potential inventory that might be available for disposal to Baita Bihor for a number of operational and postclosure scenarios and associated conceptual models. The results showed that calculated impacts were below the relevant regulatory criteria. In light of the assessment, a number of recommendations relating to repository operation, optimisation of repository engineering and waste disposals, and environmental monitoring were made. (authors)

Project Guarantee 1985. Final repository for high-level radioactive wastes: The system of safety barriers

International Nuclear Information System (INIS)

Anon.

1985-01-01

Final disposal of radioactive waste involves preventing the waste from returning from the repository location into the biosphere by means of successively arranged containment measures known as safety barriers. In the present volume NGB 85-04 of the series of reports for Project 'Guarantee' 1985, the safety barrier system for the type C repository for high-level waste is described. The barrier parameters which are relevant for safety analysis are quantified and associated error limits and data scatter are given. The aim of the report is to give a summary documentation of the safety analysis input data and their scientific background. For secure containment of radioactive waste safety barriers are used which effectively limit the release of radioactive material from the repository (release barriers) and effectively retard the entry of the original radioactive material into the biosphere (time barriers). Safety barriers take the form of both technically constructed containment measures and the siting of the repository in suitable geological formations. The technical safety barrier system in the case of high-level waste comprises: the waste solidification matrix (borosilicate glass), massive steel canisters, encasement of the waste canisters, encasement of the waste canisters in highly compacted bentonite, sealing of vacant storage space and access routes on repository closure. The natural geological safety barriers - the host rock and overlying formations provide sufficiently long deep groundwater flow times from the repository location to the earth's surface and for additional lengthening of radionuclide migration times by means of various chemical and physical retardation mechanisms. The stability of the geological formations is so great that hydrogeological system is protected for a sufficient length of time from deterioration caused, in particular, by erosion. Observations in the final section of the report indicate that input data for the type C repository safety

Evaluation of methods and tools to develop safety concepts and to demonstrate safety for an HLW repository in salt. Final report

Energy Technology Data Exchange (ETDEWEB)

Bollingerfehr, W.; Buhmann, D.; Doerr, S.; and others

2017-03-15

Salt formations have been the preferred option as host rocks for the disposal of high level radioactive waste in Germany for more than 40 years. During this period comprehensive geological investigations have been carried out together with a broad spectrum of concept and safety related R and D work. The behaviour of an HLW repository in salt formations, particularly in salt domes, has been analysed in terms of assessment of the total system performance. This was first carried out for concepts of generic waste repositories in salt and, since 1998, for a repository concept with specific boundary conditions, taking the geology of the Gorleben salt dome as an example. Suitable repository concepts and designs were developed, the technical feasibility has been proven and operational and long-term safety evaluated. Numerical modelling is an important input into the development of a comprehensive safety case for a waste repository. Significant progress in the development of numerical tools and their application for long-term safe ty assessment has been made in the last two decades. An integrated approach has been used in which the repository concept and relevant scientific and engineering data are combined with the results from iterative safety assessments to increase the clarity and the traceability of the evaluation. A safety concept that takes full credit of the favourable properties of salt formations was developed in the course of the R and D project ISIBEL, which started in 2005. This concept is based on the safe containment of radioactive waste in a specific part of the host rock formation, termed the containment providing rock zone, which comprises the geological barrier, the geotechnical barriers and the compacted backfill. The future evolution of the repository system will be analysed using a catalogue of Features, Events and Processes (FEP), scenario development and numerical analysis, all of which are adapted to suit the safety concept. Key elements of the

Evaluation of methods and tools to develop safety concepts and to demonstrate safety for an HLW repository in salt. Final report

International Nuclear Information System (INIS)

Bollingerfehr, W.; Buhmann, D.; Doerr, S.

2017-03-01

Salt formations have been the preferred option as host rocks for the disposal of high level radioactive waste in Germany for more than 40 years. During this period comprehensive geological investigations have been carried out together with a broad spectrum of concept and safety related R and D work. The behaviour of an HLW repository in salt formations, particularly in salt domes, has been analysed in terms of assessment of the total system performance. This was first carried out for concepts of generic waste repositories in salt and, since 1998, for a repository concept with specific boundary conditions, taking the geology of the Gorleben salt dome as an example. Suitable repository concepts and designs were developed, the technical feasibility has been proven and operational and long-term safety evaluated. Numerical modelling is an important input into the development of a comprehensive safety case for a waste repository. Significant progress in the development of numerical tools and their application for long-term safe ty assessment has been made in the last two decades. An integrated approach has been used in which the repository concept and relevant scientific and engineering data are combined with the results from iterative safety assessments to increase the clarity and the traceability of the evaluation. A safety concept that takes full credit of the favourable properties of salt formations was developed in the course of the R and D project ISIBEL, which started in 2005. This concept is based on the safe containment of radioactive waste in a specific part of the host rock formation, termed the containment providing rock zone, which comprises the geological barrier, the geotechnical barriers and the compacted backfill. The future evolution of the repository system will be analysed using a catalogue of Features, Events and Processes (FEP), scenario development and numerical analysis, all of which are adapted to suit the safety concept. Key elements of the

Deep repository for long-lived low- and intermediate-level waste. Preliminary safety assessment

International Nuclear Information System (INIS)

1999-11-01

A preliminary safety assessment has been performed of a deep repository for long-lived low- and intermediate-level waste, SFL 3-5. The purpose of the study is to investigate the capacity of the facility to act as a barrier to the release of radionuclides and toxic pollutants, and to shed light on the importance of the location of the repository site. A safety assessment (SR 97) of a deep repository for spent fuel has been carried out at the same time. In SR 97, three hypothetical repository sites have been selected for study. These sites exhibit fairly different conditions in terms of hydrogeology, hydrochemistry and ecosystems. To make use of information and data from the SR 97 study, we have assumed that SFL 3-5 is co-sited with the deep repository for spent fuel. A conceivable alternative is to site SFL 3-5 as a completely separate repository. The focus of the SFL 3-5 study is a quantitative analysis of the environmental impact for a reference scenario, while other scenarios are discussed and analyzed in more general terms. Migration in the repository's near- and far-field has been taken into account in the reference scenario. Environmental impact on the three sites has also been calculated. The calculations are based on an updated forecast of the waste to be disposed of in SFL 3-5. The forecast includes radionuclide content, toxic metals and other substances that have a bearing on a safety assessment. The safety assessment shows how important the site is for safety. Two factors stand out as being particularly important: the water flow at the depth in the rock where the repository is built, and the ecosystem in the areas on the ground surface where releases may take place in the future. Another conclusion is that radionuclides that are highly mobile and long-lived, such as 36 Cl and 93 Mo , are important to take into consideration. Their being long-lived means that barriers and the ecosystems must be regarded with a very long time horizon

Conclusion of the Preliminary Safety report for the LILW Repository on Trgovska Gora

International Nuclear Information System (INIS)

Lokner, V.; Levanat, I.; Schaller, A.; Kucar-Dragicevic, S.; Cerskov Klika, M.; Subasic, D.

2002-01-01

For more than a decade, APO d.o.o. has been engaged in preparations which might lead to establishment of a radioactive waste repository on Trgovska Gora, suitable for disposal of low and intermediate level waste (LILW) from the nuclear power plant Krsko. A recent product of theses activities is the preliminary safety assessment report (PSAR) for the proposed repository. In addition to an extensive overview of the repository project status, this preliminary SAR describes how the safety assessment methodology is used to demonstrate that a LILW facility will comply with radiological protection and safety requirements after the repository closure. LILW repository is designed to isolate waste from the environment for a couple hundred years in a reasonably efficient manner. It is generally not practicable to grant full waste containment throughout that period, because it suffices to demonstrate that radionuclide release and migration will remain below acceptable levels, which is achieved through safety assessment scenarios, modeling and calculations. However, with very limited repository specific data, safety assessment can only produce a conservative estimate of the upper bounds of potential exposures the repository could inflict. This PSAR arrives at such estimates in two different ways: (a) by simple bounding calculations and (b) through more sophisticated modeling and application of dedicated computer codes, but with similar conservative assumptions. Both approaches conservatively estimate that the highest potential dose to a nearby resident cannot significantly exceed the dose constraint of 0.2 mSv per year. Only in case of inadvertent intrusion into the near-surface disposal vault, much higher doses might be inflicted immediately after the planned institutional control of 250 years expires, but that can be prevented by a longer control period. Despite the preliminary and bounding style of the calculations, the PSAR has identified most important assumptions and

Nuclear safety requirements for upgrading the National Repository for Radioactive Wastes-Baita Bihor

International Nuclear Information System (INIS)

Vladescu, Gabriela; Necula, Daniela

2000-01-01

The upgrading project of National Repository for Radioactive Wastes-Baita Bihor is based on the integrated concept of nuclear safety. Its ingredients are the following: A. The principles of nuclear safety regarding the management of radioactive wastes and radioprotection; B. Safety objectives for final disposal of low- and intermediate-level radioactive wastes; C. Safety criteria for final disposal of low- and intermediate-level radioactive wastes; D. Assessment of safety criteria fulfillment for final disposal of low- and intermediate-level radioactive wastes. Concerning the nuclear safety in radioactive waste management the following issues are considered: population health protection, preventing transfrontier contamination, future generation radiation protection, national legislation, control of radioactive waste production, interplay between radioactive waste production and management, radioactive waste repository safety. The safety criteria of final disposal of low- and intermediate-level radioactive wastes are discussed by taking into account the geological and hydrogeological configuration, the physico-chemical and geochemical characteristics, the tectonics and seismicity conditions, extreme climatic potential events at the mine location. Concerning the requirements upon the repository, the following aspects are analyzed: the impact on environment, the safety system reliability, the criticality control, the filling composition to prevent radioactive leakage, the repository final sealing, the surveillance. Concerning the radioactive waste, specific criteria taken into account are the radionuclide content, the chemical composition and stability, waste material endurance to heat and radiation. The waste packaging criteria discussed are the mechanical endurance, materials toughness and types as related to deterioration caused by handling, transportation, storing or accidents. Fulfillment of safety criteria is assessed by scenarios analyses and analyses of

International comparison of safety criteria applied to radwaste repositories. Safety aspects of the post-operational phase

International Nuclear Information System (INIS)

Baltes, B.

1994-01-01

There is a generally accepted system of framework safety conditions governing the construction, operation, and post-operational monitoring of radwaste repositories. Although the development of these framework conditions may vary from country to country, the resulting criteria are based on the commonly accepted system of priciples and purposes established for ultimate radioactive waste disposal. The experience accumulated by GRS in the course of the plan approval procedure for the Konrad mine site and the safety-relevant studies performed for the planned Morsleben repository clearly show demand for further development of the safety criteria. In Germany, it is especially the safety criteria and detailed requirements filling the framework safety conditions that need revision and in-depth definition, as well as comparison and harmonisation with internationally applied criteria. These activities will particularly consider the international convention on radioactive waste management currently in preparation under the auspieces of the IAEA. (orig.) [de

Evaluation of behaviour and Safety in a geologic deep repository

International Nuclear Information System (INIS)

1997-01-01

This report presents a comprehensive description of the post-closure radiological safety assessment of a repository for the spent fuel arisings resulting from the Spanish nuclear program. This Safety Assessment constitutes a first step within a systematical process that will permit, thorough successive approximations, to predict the performance of the different barriers of the disposal system, and its capability to comply with the assigned safety functions and with the established safety criteria. The primary bases for this Safety Assessment are the following: The disposal concept considers the storage of the fuel assemblies in carbon steel canisters of 10 cm of thickness, emplaced horizontally in galleries excavated in granite of 2,4 m of diameter and 500 m of length, using a bentonite thickness of 75 cm around canisters as buffer material. The repository is located in a granitic site defined with available data about surface characteristics of Spanish granites. The exercise uses a probabilistic approximation in order to cope with the uncertainties associated with the different imputs parameters. (Author)

Deep repository for long-lived low- and intermediate-level waste. Preliminary safety assessment

Energy Technology Data Exchange (ETDEWEB)

NONE

1999-11-01

A preliminary safety assessment has been performed of a deep repository for long-lived low- and intermediate-level waste, SFL 3-5. The purpose of the study is to investigate the capacity of the facility to act as a barrier to the release of radionuclides and toxic pollutants, and to shed light on the importance of the location of the repository site. A safety assessment (SR 97) of a deep repository for spent fuel has been carried out at the same time. In SR 97, three hypothetical repository sites have been selected for study. These sites exhibit fairly different conditions in terms of hydrogeology, hydrochemistry and ecosystems. To make use of information and data from the SR 97 study, we have assumed that SFL 3-5 is co-sited with the deep repository for spent fuel. A conceivable alternative is to site SFL 3-5 as a completely separate repository. The focus of the SFL 3-5 study is a quantitative analysis of the environmental impact for a reference scenario, while other scenarios are discussed and analyzed in more general terms. Migration in the repository's near- and far-field has been taken into account in the reference scenario. Environmental impact on the three sites has also been calculated. The calculations are based on an updated forecast of the waste to be disposed of in SFL 3-5. The forecast includes radionuclide content, toxic metals and other substances that have a bearing on a safety assessment. The safety assessment shows how important the site is for safety. Two factors stand out as being particularly important: the water flow at the depth in the rock where the repository is built, and the ecosystem in the areas on the ground surface where releases may take place in the future. Another conclusion is that radionuclides that are highly mobile and long-lived, such as {sup 36}Cl and {sup 93}Mo , are important to take into consideration. Their being long-lived means that barriers and the ecosystems must be regarded with a very long time horizon.

Compliance demonstration: What can be reasonably expected from safety assessment for geological repositories?

International Nuclear Information System (INIS)

Zuidema, P.; Smith, P.; Sumerling, T.

1999-01-01

When licensing a nuclear facility, it is important to demonstrate that it will comply with regulatory limits (e.g. individual dose limits) and also show that sufficient attention has been paid to optimisation of facility design and operation, such that any associated radiological impacts will be as low as reasonably achievable (ALARA). In general, in demonstrating compliance, experience can be drawn from the performance of existing and similar facilities, and monitoring plans can be specified that will confirm that actual radiological discharges during operations are within authorised limits for the facility. This is also true in respect of the operational period of a geological repository. For the post-closure phase of a repository, however, it is also necessary to show that possible releases will remain acceptably low even at long times in the future when, it is assumed, control of the facility has lapsed and there is no method of either monitoring releases or taking remedial action in the case of unexpected events or releases. In addition, within each country, a deep geological repository will be a first-of-a-kind development so that compliance arguments can be expected to be rigorously tested without any assistance from the precedent of licensing of similar facilities nationally. This puts heavy, and quite unusual, burdens on the long-term safety assessment for a geological repository to develop a case that is sufficiently strong to demonstrate compliance. This paper focuses on the problem of demonstrating compliance with long-term safety requirements for a geological repository, and explores: the overall aims and special difficulties of demonstrating compliance for a geological repository; the role of safety assessment in demonstrating compliance; the scope for optimisation of a geological repository and importance of robustness and lessons learnt from the application of safety assessment. In addition, some issues requiring further discussion and clarification

Project Guarantee 1985. Final repository for low- and intermediate-level radioactive wastes: The system of safety barriers

International Nuclear Information System (INIS)

Anon.

1985-01-01

The safety barrier system for the type B repository for low- and intermediate-level waste is described. The barrier parameters which are relevant for safety analysis are quantified and associated error limits and data scatter are given. The aim of the report is to give a summary documentation of the safety analysis input data and their scientific background. For secure containment of radioactive waste safety barriers are used which effectively limit the release of radioactive material from the repository (release barriers) and effectively retard the entry of the original radioactive material into the biosphere (time barriers). In the case of low- and intermediate-level waste the technical safety barrier system comprises: waste solidification matrix (cement, bitumen and resin), immobilisation of the waste packages in containers using liquid cement, concrete repository containers, backfilling of remaining vacant storage space with special concrete, concrete lining of the repository caverns, sealing of access tunnels on final closure of the repository. Natural geological safety barriers - host rock and overlying formations - have the following important functions. Because of its stability, the host rock in the repository zone protects the technical safety barrier system from destruction caused by climatic effects and erosion for a sufficient length of time. It also provides for low water flow and favourable chemistry (reducing conditions)

Building the safety case for a hypothetical underground repository in crystalline rock. Final report. Vol. 2. Safety file

International Nuclear Information System (INIS)

Biurrun, E.; Engelmann, H.J.; Jobmann, M.; Lommerzheim, A.; Popp, W.; Frentz, R.R. v.; Wahl, A.

1996-10-01

The study was intended as a desk simulation of the process of preparing a licensing application for a deep repository for spent fuel and high level waste in crystalline rock. After clarifying of organizational aspects of table of contents specifying all aspects in a safety life for license application were considered. The volume II is subdivided in two parts. Part A describes the general information, waste description, site characteristics, disposal facility design, reporitory construction and operation, quality assurance, operational safety, repository closure, organization and financial aspects, and long-term safety assessment. Part B deals with the impact of retrievability. (DG)

Safety assessment of Novi Han radioactive waste repository - features, problems, results and perspectives

International Nuclear Information System (INIS)

Mateeva, M.

2000-01-01

This paper summarizes the work done and the achievements reached in the Novi Han radioactive waste repository safety assessment within the IAEA Model Project 'Increasing the safety of Novi Han radioactive waste repository BUL 4/005'. The overall safety assessment has a wide context, but the work reported here relates only to some details and results concerning the development and implementation of the appropriate methodology approach, model and computer code used for the calculations. Different steps and procedures are included for a better practical understanding of the obtained results during the safety assessment performance. The methodology approach is widely based on an international experience in safety analysis and implemented for evaluation computer code AMBER, which is one of the recommended from the safety assessments experts. (author)

Aspects of operational safety and long-term structural stability of the Morsleben repository for radioactive wastes (ERAM)

International Nuclear Information System (INIS)

Wernicke, R.S.

1997-01-01

The results of safety evaluations and safety reports reveal undoubtedly, that the Morsleben final repository operations is safe and responsible. On the basis of safety-technical evaluations some need was identified for locally optimizing the repository operations and possibly also some geotechnical features of the mine. However, this does not raise safety-related questions for man and the environment. In addition to the control exercised by the supervisory body, continuing evaluations of the repository operations assure, that changes of the safety status will be recognized in a timely manner and that competent action may be taken if necessary. (orig.) [de

Evaluating the Long-Term Safety of a Repository at Yucca Mountain

International Nuclear Information System (INIS)

Luik, Abe Van

2002-01-01

Regulations require that the repository be evaluated for its health and safety effects for 10,000 years for the Site Recommendation process. Regulations also require potential impacts to be evaluated for up to a million years in an Environmental Impact Statement. The Yucca Mountain Project is in the midst of the Site Recommendation process. The Total System Performance Assessment (TSPA) that supports the Site Recommendation evaluated safety for these required periods of time. Results showed it likely that a repository at this site could meet the licensing requirements promulgated by the Nuclear Regulatory Commission. The TSPA is the tool that integrates the results of many years of scientific investigations with design information to allow evaluations of potential far-future impacts of building a Yucca Mountain repository. Knowledge created in several branches of physics is part of the scientific basis of the TSPA that supports the Site Recommendation process.

Natural Elemental Concentrations and Fluxes: Their Use as Indicators of Repository Safety

Energy Technology Data Exchange (ETDEWEB)

Miller, Bill; Lind, Andy; Savage, Dave; Maul, Philip; Robinson, Peter [EnvirosQuantisci, Melton Mowbray (United Kingdom)

2002-03-01

The calculated post-closure performance of a radioactive waste repository is generally quantified in terms of radiological dose or risk to humans, with safety being determined by whether the calculated exposure values are consistent with predetermined target criteria which are deemed to represent acceptable radiological hazards. Despite their general acceptance, however, dose and risk are not perfect measures of repository safety because, in order to calculate them, gross assumptions must be made for future human behaviour patterns. Such predictions clearly become increasingly uncertain as forecasts are made further into the future. As a consequence, there has been a growing interest in developing other ways of assessing repository safety which do not require assumptions to be made for future human behaviour. One proposed assessment method is to use the distributions of naturally-occurring chemical species in the environment, expressed either as concentrations or fluxes of elements, radionuclides or radioactivity, as natural safety indicators which may be compared with the PA predictions of repository releases. Numerous comparisons are possible between the repository and natural systems. The primary objective is to use the natural system to provide context to the hazard presented by the repository releases. Put simply, if it can be demonstrated that the flux to the biosphere from the repository is not significant compared with the natural flux from the geosphere, then its radiological significance should not be of great or priority concern. Natural safety indicators may be quantified on a site specific basis, using information derived from a repository site characterisation programme, and can be compared to the outputs from the associated site specific PAs. Such calculations and comparisons may be very detailed and might examine, for example, the spatial and temporal variations in the distributions and fluxes of naturally-occurring chemical species arising from

The use of uranium fluxes as safety indicators of radioactive waste repositories

International Nuclear Information System (INIS)

Miller, W.M.; Hooker, P.J.

2002-01-01

Natural analogues based on uranium deposits are commonly used to represent the long-term behaviour of radioactive waste repositories or the processes that influence their radioactive contents. The geochemical dispersion of naturally occurring uranium can also be used to model natural radioactivity fluxes in the vicinity of a planned repository. These fluxes can be estimated for erosional and groundwater discharge processes and compared with calculated future fluxes of radioactivity that would be released from a repository. The methodology is outlined and the benefits of the approach for supporting the derivation of a safety case for a repository are indicated. (author)

Safety considerations of disposal of disused sealed sources in Puspokszilagy Repository, Hungary

International Nuclear Information System (INIS)

2003-01-01

The report presents the management of radioactive waste in Hungary Puspokszilagy Repository (RWTDF) including waste acceptance criteria, safety assessments, Action Plan for the safety improvement and present projects. The Puspokszilagy Repository is a typical near-surface repository, sink into the ground 6 m depth. The facility is a shallow land disposal type, appropriated for disposal of short and medium lived LILW, acceptable for temporary storage of long lived LILW. It consists of vaults containing cells for solidified drummed waste, wells for spent sealed sources, work building for treatment and interim storage and office building for environmental measurements. Two safety assessments have been performed in 2000 and 2002. The new safety assessment confirms the main statements of SA 2000, according to which several waste types can cause serious problems in the distant future: Until the finish of passive control the safety of the environment is guaranteed. After that time it is possible to arise events leading to exceeding of dose restricts (more then 10 mSv/yr but less then 100 mSv/yr), because of disposal of long lived radionuclides (mainly C-14,Tc-99, Ra-226, Th-232, U-234) and significant activities of Cs-137 sources.There are uncertainties in radionuclide amounts and distributions, as well as in the physical and chemical characteristics of the wastes that determine radionuclide mobility and toxicity. The recommendations to improve the safety include: Long lived SSRS in the 'B' and 'D' wells should be removed before the closure of repository. Large Cs-137 sources and long lived sources in the 'A' vaults should be recovered (if its feasible); All vaults should be backfilled to provide chemical conditioning; The waste packaged in plastic bags should be repackaged and compacted into drums or containers; The inventory should be revise. Waste acceptance requirements in the future are: The disposal of long lived radionuclides is no permitted. The long lived waste

Natural Elemental Concentrations and Fluxes: Their Use as Indicators of Repository Safety

International Nuclear Information System (INIS)

Miller, Bill; Lind, Andy; Savage, Dave; Maul, Philip; Robinson, Peter

2002-03-01

The calculated post-closure performance of a radioactive waste repository is generally quantified in terms of radiological dose or risk to humans, with safety being determined by whether the calculated exposure values are consistent with predetermined target criteria which are deemed to represent acceptable radiological hazards. Despite their general acceptance, however, dose and risk are not perfect measures of repository safety because, in order to calculate them, gross assumptions must be made for future human behaviour patterns. Such predictions clearly become increasingly uncertain as forecasts are made further into the future. As a consequence, there has been a growing interest in developing other ways of assessing repository safety which do not require assumptions to be made for future human behaviour. One proposed assessment method is to use the distributions of naturally-occurring chemical species in the environment, expressed either as concentrations or fluxes of elements, radionuclides or radioactivity, as natural safety indicators which may be compared with the PA predictions of repository releases. Numerous comparisons are possible between the repository and natural systems. The primary objective is to use the natural system to provide context to the hazard presented by the repository releases. Put simply, if it can be demonstrated that the flux to the biosphere from the repository is not significant compared with the natural flux from the geosphere, then its radiological significance should not be of great or priority concern. Natural safety indicators may be quantified on a site specific basis, using information derived from a repository site characterisation programme, and can be compared to the outputs from the associated site specific PAs. Such calculations and comparisons may be very detailed and might examine, for example, the spatial and temporal variations in the distributions and fluxes of naturally-occurring chemical species arising from

Project Guarantee 1985. Final repository for low- and intermediate level radioactive wastes: Safety report

International Nuclear Information System (INIS)

Anon.

1985-01-01

Storage of radioactive waste must delay the return of radionuclides to the biosphere for a long period of time and must maintain the release rates at a sufficiently low level for all time. This is achieved with the aid of a series of safety barriers which consist, on the one hand, of technical barriers in the repository and, on the other hand , of natural geological barriers as they occur at the repository location. In order to assess the efficiency of the barriers, the working methods of the technical barriers and the host rock must be understood. This understanding is transferred into quantitative models in order to calculate the safety of the repository. The individual barriers and the methods used to modelling their functions were described in volume NGB 85-07 of the Project Guarantee 1985 report series and the data necessary for modelling were given. The models and data are used in the safety analysis, the results of which are contained in the present report. Safety considerations show that models are available in Switzerland which allow, in principle, an assessment of the long-term behaviour of a repository for low- and intermediate-level waste. The evaluation of earlier studies and experimental work, suitable laboratory measurements and results from field research enable compilation of a representative data-set so that the requirements for quantitative statements on safety of final disposal are met from this side also. The safety calculations show that the radiation doses calculated for a base case scenario with realistic/conservative parameter values are negligibly low. Also, radiation doses which are clearly under the protection standard of 10 mrem per year result for conservative values and the cumulation of several conservative assumptions. Even assuming exposure of the repository by erosion, a radiotoxicity of the soil formed results which is under natural values

Technical Standards on the Safety Assessment of a HLW Repository in Other Countries

International Nuclear Information System (INIS)

Lee, Sung Ho; Hwang, Yong Soo

2009-01-01

The basic function of HLW disposal system is to prevent excessive radio-nuclides being leaked from the repository in a short time. To do this, many technical standards should be developed and established on the components of disposal system. Safety assessment of a repository is considered as one of technical standards, because it produces quantitative results of the future evolution of a repository based on a reasonably simplified model. In this paper, we investigated other countries' regulations related to safely assessment focused on the assessment period, radiation dose limits and uncertainties of the assessment. Especially, in the investigation process of the USA regulations, the USA regulatory bodies' approach to assessment period and peak dose is worth taking into account in case of a conflict between peak dose from safety assessment and limited value in regulation.

Safety assessment methodology for waste repositories in deep geological formations

International Nuclear Information System (INIS)

Chapuis, A.M.; Lewi, J.; Pradel, J.; Queniart, D.; Raimbault, P.; Assouline, M.

1986-06-01

The long term safety of a nuclear waste repository relies on the evaluation of the doses which could be transferred to man in the future. This implies a detailed knowledge of the medium where the waste will be confined, the identification of the basic phenomena which govern the migration of the radionuclides and the investigation of all possible scenarios that may affect the integrity of the barriers between the waste and the biosphere. Inside the Institute of protection and nuclear safety of the French Atomic Energy Commission (CEA/IPSN), the Department of the Safety Analysis (DAS) is currently developing a methodology for assessing the safety of future geological waste repositories, and is in charge of the modelling development, while the Department of Technical Protection (DPT) is in charge of the geological experimental studies. Both aspects of this program are presented. The methodology for risk assessment stresses the needs for coordination between data acquisition and model development which should result in the obtention of an efficient tool for safety evaluation. Progress needs to be made in source and geosphere modelling. Much more sophisticated models could be used than the ones which is described; however sensitivity analysis will determine the level of sophistication which is necessary to implement. Participation to international validation programs are also very important for gaining confidence in the approaches which have been chosen

TURVA-2012 safety case for licensing a spent fuel repository at Olkiluoto, Finland

International Nuclear Information System (INIS)

Vira, Juhani; Snellman, Margit

2014-01-01

In 2001, the Finnish Parliament endorsed a decision-in-principle (DiP) whereby the spent nuclear fuel produced by the operating nuclear reactors at Olkiluoto and Loviisa will be disposed of in a geological repository at Olkiluoto, on the south-western coast of Finland. Subsequently, additional DiPs were issued allowing the extension of the repository to accommodate spent nuclear fuel from additional reactors that are under construction or in planning at Olkiluoto, which means a total of 9 000 tU of spent nuclear fuel to be disposed of. In accordance with the decision of the Ministry of Trade and Industry (KTM) in 2003, Posiva submitted an application for a license to construct a disposal facility at Olkiluoto in 2012, consisting of an encapsulation facility and an underground deep geological repository. The application included a Preliminary Safety Analysis Report (PSAR) and a long-term safety case, TURVA-2012. Assuming a positive outcome of the current licensing review, the next step would be the Final Safety Analysis Report (FSAR) in support of an operational licence application around 2020. The disposal method is based on the same KBS-3 concept that the Swedish SKB has used as basis for their license application in 2010. Accordingly, the spent nuclear fuel will be encapsulated in water- and gas-tight copper canisters equipped with a load-bearing insert and emplaced in a deep geological repository constructed in the bedrock. The canisters will be surrounded by a swelling clay buffer material that isolates them from the bedrock. The deposition tunnels and the central tunnels and the other underground openings will be backfilled with materials of low permeability. The repository will be at a depth of about 400-450 m below ground. The primary role of the bedrock is to provide sufficiently stable conditions for the engineered barrier system and to make inadvertent human intrusion unlikely. In case of EBS failure, the bedrock shall also retain and retard the possible

Modeling and Analysis on Radiological Safety Assessment of Low- and Intermediate Level Radioactive Waste Repository

Energy Technology Data Exchange (ETDEWEB)

Lee, Youn Myoung; Jung, Jong Tae; Kang, Chul Hyung (and others)

2008-04-15

Modeling study and analysis for technical support for the safety and performance assessment of the low- and intermediate level (LILW) repository partially needed for radiological environmental impact reporting which is essential for the licenses for construction and operation of LILW has been fulfilled. Throughout this study such essential area for technical support for safety and performance assessment of the LILW repository and its licensing as gas generation and migration in and around the repository, risk analysis and environmental impact during transportation of LILW, biosphere modeling and assessment for the flux-to-dose conversion factors for human exposure as well as regional and global groundwater modeling and analysis has been carried out.

Modeling and Analysis on Radiological Safety Assessment of Low- and Intermediate Level Radioactive Waste Repository

International Nuclear Information System (INIS)

Lee, Youn Myoung; Jung, Jong Tae; Kang, Chul Hyung

2008-04-01

Modeling study and analysis for technical support for the safety and performance assessment of the low- and intermediate level (LILW) repository partially needed for radiological environmental impact reporting which is essential for the licenses for construction and operation of LILW has been fulfilled. Throughout this study such essential area for technical support for safety and performance assessment of the LILW repository and its licensing as gas generation and migration in and around the repository, risk analysis and environmental impact during transportation of LILW, biosphere modeling and assessment for the flux-to-dose conversion factors for human exposure as well as regional and global groundwater modeling and analysis has been carried out

A Deterministic Safety Assessment of a Pyro-processed Waste Repository

International Nuclear Information System (INIS)

Lee, Youn Myoung; Jeong, Jong Tae; Choi, Jong Won

2012-01-01

A GoldSim template program for a safety assessment of a hybrid-typed repository system, called 'A-KRS', in which two kinds of pyro-processed radioactive wastes, low-level metal wastes and ceramic high-level wastes that arise from the pyro-processing of PWR nuclear spent fuels are disposed of, has been developed. This program is ready both for a deterministic and probabilistic total system performance assessment which is able to evaluate nuclide release from the repository and farther transport into the geosphere and biosphere under various normal, disruptive natural and manmade events, and scenarios. The A-KRS has been deterministically assessed with 5 various normal and abnormal scenarios associated with nuclide release and transport in and around the repository. Dose exposure rates to the farming exposure group have been evaluated in accordance with all the scenarios and then compared among other.

1972 preliminary safety analysis report based on a conceptual design of a proposed repository in Kansas

International Nuclear Information System (INIS)

Blomeke, J.O.

1977-08-01

This preliminary safety analysis report is based on a proposed Federal Repository at Lyons, Kansas, for receiving, handling, and depositing radioactive solid wastes in bedded salt during the remainder of this century. The safety analysis applies to a hypothetical site in central Kansas identical to the Lyons site, except that it is free of nearby salt solution-mining operations and bore holes that cannot be plugged to Repository specifications. This PSAR contains much information that also appears in the conceptual design report. Much of the geological-hydrological information was gathered in the Lyons area. This report is organized in 16 sections: considerations leading to the proposed Repository, design requirements and criteria, a description of the Lyons site and its environs, land improvements, support facilities, utilities, different impacts of Repository operations, safety analysis, design confirmation program, operational management, requirements for eventually decommissioning the facility, design criteria for protection from severe natural events, and the proposed program of experimental investigations

1972 preliminary safety analysis report based on a conceptual design of a proposed repository in Kansas

Energy Technology Data Exchange (ETDEWEB)

Blomeke, J.O.

1977-08-01

This preliminary safety analysis report is based on a proposed Federal Repository at Lyons, Kansas, for receiving, handling, and depositing radioactive solid wastes in bedded salt during the remainder of this century. The safety analysis applies to a hypothetical site in central Kansas identical to the Lyons site, except that it is free of nearby salt solution-mining operations and bore holes that cannot be plugged to Repository specifications. This PSAR contains much information that also appears in the conceptual design report. Much of the geological-hydrological information was gathered in the Lyons area. This report is organized in 16 sections: considerations leading to the proposed Repository, design requirements and criteria, a description of the Lyons site and its environs, land improvements, support facilities, utilities, different impacts of Repository operations, safety analysis, design confirmation program, operational management, requirements for eventually decommissioning the facility, design criteria for protection from severe natural events, and the proposed program of experimental investigations. (DLC)

Evaluation of radiological safety assessment of a repository in a clay rock formation

International Nuclear Information System (INIS)

1999-01-01

This report presents a comprehensive description of the post-closure radiological safety assessment of a repository for the spent fuel arisings resulting from the Spanish nuclear program excavated in a clay host rock formation. In this report three scenarios have been analysed in detail. The first scenario represents the normal in detail. The first scenario represents the normal evolution of the repository (Reference Scenario); and includes a set of variants to investigate the relative importance of the various repository components and examine the sensitivity of the performance to parameters variations. Two altered scenarios have also been considered: deep well construction and poor sealing of the repository. This document contains a detailed description of the repository system, the methodology adopted for the scenarios generation, the process modelling approach and the results of the consequences analysis. (Author)

Review. Deep repository for spent nuclear fuel SR 97 - Post-closure safety

International Nuclear Information System (INIS)

Stephansson, Ove

2000-01-01

SKB states that the chosen scenarios provide good coverage of future evolutionary pathways for the deep repository. This is not the case. SKB has not made full use of the established interaction matrices and the new method of THMC diagrams to generate the relevant and important scenarios and to construct the important pathways of variables and processes, either in the established interaction matrices and the presented THMC diagrams. Hence, SKB is demonstrating in SR 97 that they lack a well thought through, sound and solid method to select and evaluate scenarios for the purpose of demonstrating the safety of a deep repository for spent nuclear fuel. The evolution of the system is presented for the components of the repository system (fuel, canister, buffer/backfill, geosphere) and the effects of four different scenarios, but time only enters into the system for discrete events or processes, e.g. description of the relative radiotoxicity and heat decay of the fuel, temperature distribution, iron exchange process, pH in buffer, redox capacity and radionuclear release at the three sites. There is a lack of method and way of describing the evolution of the complete repository system, including the major scenarios, as a function of time. It is essential that SKB is able to: - consider the full range of potential scenarios, - grade the scenarios according to their significance for repository design and performance and safety assessment, - consider whether simple engineering actions could be taken to inhibit the development of adverse scenarios. This cannot be done with the system presented in SR 97, and so SKB do not have a full predictive capability - which is required for the engineering design of such an important and costly structure as a repository. Geoscientific investigation material for three selected sites are presented by SKB in the technical report dealing with waste, repository design and sites. Here a general overview is missing of the geological and rock

Preliminary post-closure safety assessment of repository concepts for low level radioactive waste at the Bruce Site, Ontario

International Nuclear Information System (INIS)

Little, R.H.; Penfold, J.S.S.; Egan, M.J.; Leung, H.

2005-01-01

The preliminary post-closure safety assessment of permanent repository concepts for low-level radioactive waste (LLW) at the Ontario Power Generation (OPG) Bruce Site is described. The study considered the disposal of both short and long-lived LLW. Four geotechnically feasible repository concepts were considered (two near-surface and two deep repositories). An approach consistent with best international practice was used to provide a reasoned and comprehensive analysis of post-closure impacts of the repository concepts. The results demonstrated that the deep repository concepts in shale and in limestone, and the surface repository concept on sand should meet radiological protection criteria. For the surface repository concept on glacial till, it appears that increased engineering such as grouting of waste and voids should be considered to meet the relevant dose constraint. Should the project to develop a permanent repository for LLW proceed, it is expected that this preliminary safety assessment would need to be updated to take account of future site-specific investigations and design updates. (author)

Developing design premises for a KBS-3V repository based on results from the safety assessment - 16027

International Nuclear Information System (INIS)

Andersson, Johan; Hedin, Allan

2009-01-01

As a part of the planned license application for a final repository for spent nuclear fuel the Swedish Nuclear Fuel and Waste Management Co. (SKB), has developed design premises from a long term safety aspect of a KBS-3V repository for spent nuclear fuel. The purpose is to provide requirements from a long term safety aspect, to form the basis for the development of the reference design of the repository and to justify that design. Design premises typically concern specification on what mechanical loads the barriers must withstand, restrictions on the composition of barrier materials or acceptance criteria for the various underground excavations. These design constraints, if all fulfilled by the actual design, should form a good basis for demonstrating repository safety. The justification for these design premises is derived from SKB's most recent safety assessment SR-Can complemented by a few additional analyses. Some of the design premises may be modified in future stages of SKB's program, as a result of analyses based on more detailed site data and a more developed understanding of processes of importance for long-term safety. (authors)

A Generic Safety Assessment Model for a Trench Type LILW Repository

International Nuclear Information System (INIS)

Lee, Youn-Myoung; Choi, Hee-Joo

2015-01-01

This program is ready for a total system performance assessment and is able to deterministically and probabilistically evaluate the nuclide release from a repository and farther transport into the geosphere and biosphere under various normal circumstances, disruptive events, and scenarios that can occur after a failure of waste packages with associated uncertainty. Despite the conceptual design of a trench type LILW repository system, all parameter values associated with the repository system were assumed for the time being, and the generic model developed through this study should be helpful because the evaluation of such releases is very important. A simple and effective model for a safety assessment of a conceptual trench repository system, in which an LILW that arises from a nuclear power plant and other sources, has been developed. The computer program based on this model has also been developed as a GoldSim template using the commercial GoldSim development tool

A Generic Safety Assessment Model for a Trench Type LILW Repository

Energy Technology Data Exchange (ETDEWEB)

Lee, Youn-Myoung; Choi, Hee-Joo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-05-15

This program is ready for a total system performance assessment and is able to deterministically and probabilistically evaluate the nuclide release from a repository and farther transport into the geosphere and biosphere under various normal circumstances, disruptive events, and scenarios that can occur after a failure of waste packages with associated uncertainty. Despite the conceptual design of a trench type LILW repository system, all parameter values associated with the repository system were assumed for the time being, and the generic model developed through this study should be helpful because the evaluation of such releases is very important. A simple and effective model for a safety assessment of a conceptual trench repository system, in which an LILW that arises from a nuclear power plant and other sources, has been developed. The computer program based on this model has also been developed as a GoldSim template using the commercial GoldSim development tool.

Review. Deep repository for spent nuclear fuel SR 97 - Post-closure safety

Energy Technology Data Exchange (ETDEWEB)

Stephansson, Ove [Royal Inst. of Tech., Stockholm (Sweden). Dept. of Civil and Environmental Engineering

2000-12-01

SKB states that the chosen scenarios provide good coverage of future evolutionary pathways for the deep repository. This is not the case. SKB has not made full use of the established interaction matrices and the new method of THMC diagrams to generate the relevant and important scenarios and to construct the important pathways of variables and processes, either in the established interaction matrices and the presented THMC diagrams. Hence, SKB is demonstrating in SR 97 that they lack a well thought through, sound and solid method to select and evaluate scenarios for the purpose of demonstrating the safety of a deep repository for spent nuclear fuel. The evolution of the system is presented for the components of the repository system (fuel, canister, buffer/backfill, geosphere) and the effects of four different scenarios, but time only enters into the system for discrete events or processes, e.g. description of the relative radiotoxicity and heat decay of the fuel, temperature distribution, iron exchange process, pH in buffer, redox capacity and radionuclear release at the three sites. There is a lack of method and way of describing the evolution of the complete repository system, including the major scenarios, as a function of time. It is essential that SKB is able to: - consider the full range of potential scenarios, - grade the scenarios according to their significance for repository design and performance and safety assessment, - consider whether simple engineering actions could be taken to inhibit the development of adverse scenarios. This cannot be done with the system presented in SR 97, and so SKB do not have a full predictive capability - which is required for the engineering design of such an important and costly structure as a repository. Geoscientific investigation material for three selected sites are presented by SKB in the technical report dealing with waste, repository design and sites. Here a general overview is missing of the geological and rock

Scenario Analysis for the Safety Assessment of Nuclear Waste Repositories: A Critical Review.

Science.gov (United States)

Tosoni, Edoardo; Salo, Ahti; Zio, Enrico

2018-04-01

A major challenge in scenario analysis for the safety assessment of nuclear waste repositories pertains to the comprehensiveness of the set of scenarios selected for assessing the safety of the repository. Motivated by this challenge, we discuss the aspects of scenario analysis relevant to comprehensiveness. Specifically, we note that (1) it is necessary to make it clear why scenarios usually focus on a restricted set of features, events, and processes; (2) there is not yet consensus on the interpretation of comprehensiveness for guiding the generation of scenarios; and (3) there is a need for sound approaches to the treatment of epistemic uncertainties. © 2017 Society for Risk Analysis.

Development of database systems for safety of repositories for disposal of radioactive wastes

Energy Technology Data Exchange (ETDEWEB)

Lee, Yeong Hun; Han, Jeong Sang; Shin, Hyeon Jun; Ham, Sang Won; Kim, Hye Seong [Yonsei Univ., Seoul (Korea, Republic of)

1999-03-15

In the study, GSIS os developed for the maximizing effectiveness of the database system. For this purpose, the spatial relation of data from various fields that are constructed in the database which was developed for the site selection and management of repository for radioactive waste disposal. By constructing the integration system that can link attribute and spatial data, it is possible to evaluate the safety of repository effectively and economically. The suitability of integrating database and GSIS is examined by constructing the database in the test district where the site characteristics are similar to that of repository for radioactive waste disposal.

Safety and performance assessment of geologic disposal systems for nuclear wastes

International Nuclear Information System (INIS)

Peltonen, E.

1987-01-01

This thesis presents a methodology for the safety and performance assesment of final disposal of nuclear wastes into crystalline bedrock. The applicability of radiation protection objectives is discussed, as well as the goals of the assessment in the various repository system development phases. Due consideration is given to the description of the pertinent analysis methods and to the comprehensive model system. The methodology has been applied to assess the acceptability of the basic disposal concepts and to study the possibilities for the optimization of protection. Furthermore, performance of different components in the multiple barrier disposal systems is estimated. The waste types dealt with are low- and intermediate-level waste as well as high-level spent nuclear fuel from a nuclear power plant. In addition, an option of high-level vitrified waste from reprocessing of spent fuel is taken into account. On the basis of the various analyses carried out it can be concluded that the disposal of different nuclear wastes in the Finnish bedrock in properly designed repositories meets the radiation protection objectives with good confidence. In addition, the studies indicate that the safety margins are considerable. This is due to the fact that the overall performance of the multiple barrier disposal systems analysed is not sensitive to possible unfavourable changes in barrier properties. From the optimization of protection point of view it can be concluded that there is no need to develop more effective repository designs than those analysed in this thesis. In fact, the results indicate that the most sophisticated designs have already gone beyond an optimal level of safety

Findings by the Commission Evaluating Nuclear Safety and Repository Research in Germany

International Nuclear Information System (INIS)

Sandtner, W.; Closs, K.D.

2000-01-01

The Commission Evaluating Nuclear Safety and Repository Research in Germany, which had been appointed by the German Federal Ministry of Economics on September 24, 1999, submitted its report. Here is the gist of the Commission's findings: Irrespective of the criteria established with the political decision to terminate the use of nuclear power in Germany, competence in nuclear safety must be maintained over the next few decades. Only in this way can the government perform its duty and make provisions for the future, and can the safety of nuclear facilities and waste management pathways be ensured in accordance with the international state of the art. In view of the considerable reduction in funding in recent years and also in future, measures must be taken to ensure that further decreases in-roject funding and institutionalized government financing are excluded so as to avoid further declines in terms of manpower and competence in this field. Reactor safety and repository research must be financed at a level allowing the federal government to discharge its legal duties. The full report by the Commission, with its annexes, is available on the GRS web site (http://www.grs.de) as a PDF file. (orig.) [de

Safe disposal of radioactive waste. Post-closure safety assessment of permanent repository in Novi han

International Nuclear Information System (INIS)

Mateeva, M.

2007-01-01

A presented material is the third part of the monograph with title 'Safe disposal of radioactive waste. Post-closure safety assessment of the permanent repository in Novi Han'. This part deals with review of the scenario selection procedure. The process system of permanent repository for radioactive waste is describing in details for different levels. Preliminary screening process of features, events and processes is presented here. Interaction matrixes for basic disposal system components are constructed. Final selection and grouping between the included features, events and processes is done. Selected and defined scenarios for post-closure safety assessment are presented too. Key words: post-closure safety assessment, scenario generation procedure, process system, process influence diagram, and interaction matrix

Workshop on Regulatory Review and Safety Assessment Issues in Repository Licensing

International Nuclear Information System (INIS)

Wilmot, Roger D.

2011-02-01

The workshop described here was organised to address more general issues regarding regulatory review of SKB's safety assessment and overall review strategy. The objectives of the workshop were: - to learn from other programmes' experiences on planning and review of a license application for a nuclear waste repository, - to offer newly employed SSM staff an opportunity to learn more about selected safety assessment issues, and - to identify and document recommendations and ideas for SSM's further planning of the licensing review

Integrated safety case development for deep geological repositories

International Nuclear Information System (INIS)

Kawamura, Hideki; McKinley, Ian G.

2008-01-01

The paper will illustrate an 'integrated safety case', which involves combining both pre-closure and post-closure safety arguments from the point of view of a repository implementer, who must also ensure that projects are practical, acceptable and economic. The post-closure safety case is based on the performance of a number of barriers, which are established during construction, operation and closure. Such barriers must be confirmed using quality assured methods, supported, as required, by inspection and monitoring. The requirement for integrated assessment means that even the final process to end institutional control and transfer any liabilities from the implementer needs to be considered at present, even though this will undoubtedly be refined and tailored to the site characteristics over the many decades that will pass before this occurs. To illustrate the practical application of this approach, assessment of variants for remote-handled emplacement of the EBS for disposal of HLW in Japan will be discussed. (author)

Long-term safety for the final repository for spent nuclear fuel at Forsmark. Main report of the SR-Site project

Energy Technology Data Exchange (ETDEWEB)

2011-03-15

The central conclusion of the safety assessment SR-Site is that a KBS-3 repository that fulfils long-term safety requirements can be built at the Forsmark site. This conclusion is reached because the favourable properties of the Forsmark site ensure the required long-term durability of the barriers of the KBS-3 repository. In particular, the copper canisters with their cast iron inserts have been demonstrated to provide a sufficient resistance to the mechanical and chemical loads to which they may be subjected in the repository environment. The conclusion is underpinned by: - The reliance of the KBS-3 repository on i) a geological environment that exhibits long-term stability with respect to properties of importance for long-term safety, i.e. mechanical stability, low groundwater flow rates at repository depth and the absence of high concentrations of detrimental components in the groundwater, and ii) the choice of naturally occurring materials (copper and bentonite clay) for the engineered barriers that are sufficiently durable in the repository environment to provide the barrier longevity required for safety. - The understanding, through decades of research at SKB and in international collaboration, of the phenomena that affect long-term safety, resulting in a mature knowledge base for the safety assessment. - The understanding of the characteristics of the site through several years of surface-based investigations of the conditions at depth and of scientific interpretation of the data emerging from the investigations, resulting in a mature model of the site, adequate for use in the safety assessment. - The detailed specifications of the engineered parts of the repository and the demonstration of how components fulfilling the specifications are to be produced in a quality assured manner, thereby providing a quality assured initial state for the safety assessment. The detailed analyses demonstrate that canister failures in a one million year perspective are rare

Long-term safety for the final repository for spent nuclear fuel at Forsmark. Main report of the SR-Site project

International Nuclear Information System (INIS)

2011-03-01

The central conclusion of the safety assessment SR-Site is that a KBS-3 repository that fulfils long-term safety requirements can be built at the Forsmark site. This conclusion is reached because the favourable properties of the Forsmark site ensure the required long-term durability of the barriers of the KBS-3 repository. In particular, the copper canisters with their cast iron inserts have been demonstrated to provide a sufficient resistance to the mechanical and chemical loads to which they may be subjected in the repository environment. The conclusion is underpinned by: - The reliance of the KBS-3 repository on i) a geological environment that exhibits long-term stability with respect to properties of importance for long-term safety, i.e. mechanical stability, low groundwater flow rates at repository depth and the absence of high concentrations of detrimental components in the groundwater, and ii) the choice of naturally occurring materials (copper and bentonite clay) for the engineered barriers that are sufficiently durable in the repository environment to provide the barrier longevity required for safety. - The understanding, through decades of research at SKB and in international collaboration, of the phenomena that affect long-term safety, resulting in a mature knowledge base for the safety assessment. - The understanding of the characteristics of the site through several years of surface-based investigations of the conditions at depth and of scientific interpretation of the data emerging from the investigations, resulting in a mature model of the site, adequate for use in the safety assessment. - The detailed specifications of the engineered parts of the repository and the demonstration of how components fulfilling the specifications are to be produced in a quality assured manner, thereby providing a quality assured initial state for the safety assessment. The detailed analyses demonstrate that canister failures in a one million year perspective are rare

Mathematical simulation for safety assessment of nuclear waste repositories

International Nuclear Information System (INIS)

Brandstetter, A.; Raymond, J.R.; Benson, G.L.

1979-01-01

Mathematical models are being developed as part of the Waste Isolation Safety Assessment Program (WISAP) for assessing the post-closure safety of nuclear waste storage in geologic formations. The objective of this program is to develop the methods and data necessary to determine potential events that might disrupt the integrity of a waste repository and provide pathways for radionuclides to reach the bioshpere, primarily through groundwater transport. Four categories of mathematical models are being developed to assist in the analysis of potential release scenarios and consequences: (1) release scenario analysis models; (2) groundwater flow models; (3) contaminant transport models; and (4) radiation dose models. The development of the release scenario models is in a preliminary stage; the last three categories of models are fully operational. The release scenario models determine the bounds of potential future hydrogeologic changes, including potentially disruptive events. The groundwater flow and contaminant transport models compute the flowpaths, travel times, and concentrations of radionuclides that might migrate from a repository in the event of a breach and potentially reach the biosphere. The dose models compute the radiation doses to future populations. Reference site analyses are in progress to test the models for application to different geologies, including salt domes, bedded salt, and basalt

A Conceptual Modeling for a GoldSim Program for Safety Assessment of an LILW Repository

International Nuclear Information System (INIS)

Lee, Youn Myoung; Hwang, Yong Soo; Kang, Chul Hyung; Lee, Sung Ho

2009-12-01

Modeling study and development of a total system performance assessment (TSPA) program, by which an assessment of safety and performance for a low- and intermediate-level radioactive waste disposal repository with normal or abnormal nuclide release cases associated with the various FEPs involved in the performance of the proposed repository could be made has been carrying out by utilizing GoldSim under contract with KRMC. The report deals with a detailed conceptual modeling scheme by which a GoldSim program modules, all of which are integrated into a TSPA program as well as the input data set currently available. In-depth system models that are conceptually and rather practically described and then ready for implementing into a GoldSim program are introduced with plenty of illustrative conceptual models and sketches. The GoldSim program that will be finally developed through this project is expected to be successfully applied to the post closure safety assessment required both for the LILW repository and pyro processed repository by the regulatory body with both increased practicality and much reduced uncertainty

A Probabilistic Safety Assessment of a Pyro-processed Waste Repository

International Nuclear Information System (INIS)

Lee, Youn Myoung; Jeong, Jong Tae

2012-01-01

A GoldSim template program for a safety assessment of a hybrid-typed repository system, called A-KRS, in which two kinds of pyro-processed radioactive wastes, low-level metal wastes and ceramic high-level wastes that arise from the pyro-processing of PWR nuclear spent fuels are disposed of, has been developed. This program is ready both for a deterministic and probabilistic total system performance assessment which is able to evaluate nuclide release from the repository and farther transport into the geosphere and biosphere under various normal, disruptive natural and manmade events, and scenarios. The A-KRS has been probabilistically assessed with 9 selected input parameters, each of which has its own statistical distribution for a normal release and transport scenario associated with nuclide release and transport in and around the repository. Probabilistic dose exposure rates to the farming exposure group have been evaluated. A sensitivity of 9 selected parameters to the result has also been investigated to see which parameter is more sensitive and important to the exposure rates.

Strategy for safety case development: impact of a volunteering approach to siting a japanese HLW repository

International Nuclear Information System (INIS)

Kitayama, K.; Ishiguro, K.; Takeuchi, M.; Tsuchi, H.; Kato, T.; Sakabe, Y.; Wakasugi, K.

2008-01-01

NUMO strategy for safety case development is constrained by a staged siting approach, which has been initiated by a call for volunteer municipalities to host the HLW repository. For each site, the safety case is an important factor to be considered at the selection steps which narrow down towards the preferred repository location. This is particularly challenging, however, as every site requires a tailored repository concept, with associated performance assessment and an individual site evaluation programme all of which evolve with gradually increasing understanding of the host environment. In order to maintain flexibility without losing focus, NUMO has developed a formalized tailoring procedure, termed the NUMO Structured Approach (NSA). The NSA guides the interaction of the key site characterisation, repository design and performance assessment groups and is facilitated by tools to help the decision making associated with the tailoring process (e.g. a requirements management system) and with comparison of siting and design options (e.g. multi-attribute analysis). Pragmatically, the post-closure safety case will initially emphasize near-field processes and a robust engineering barrier system, considering the limited geological information at early stages. This will be complemented by a more realistic assessment of total system performance, as needed to compare options. In addition, efforts to rigorously assess operational phase safety and the practicality of assuring quality of the constructed engineered barriers are components of the total safety case which are receiving particular attention now, as they may better discriminate between sites while information is still limited. (authors)

THEREDA - a contribution to long-term safety of repositories of nuclear and non-nuclear wastes

International Nuclear Information System (INIS)

Altmaier, M.; Kienzler, B.; Marquardt, C.M.; Neck, V.; Voigt, W.; Wilhelm, S.

2008-01-01

Long-term safety analyses of German repositories of radioactive waste as well as underground repositories for chemical toxic waste and other uses (contaminated site remediation) urgently require a standardized, comprehensive thermodynamic reference database. The former 'Thermodynamic Standard Database Working Party' was set up to establish such a database. The activities of that group have been supported within the integrated 'THEREDA' (Thermodynamic Reference Database) project since July 2006 for an initial period of 3 years by the German Federal Ministries of Education and Research, of Economics, and by the Federal Office of Radiation Protection. THEREDA at present is composed of 5 partner institutions essentially representing the key German research institutions in the field of repository safety research. THEREDA is to improve the transparency and validity of safety analyses in Germany and, for the first time, provides consistent thermodynamic datasets for the repository options discussed in Germany. Quality levels are indicated for each thermodynamic quantity on the basis of unambiguously defined evaluation criteria, which allow users to either include or exclude data in accordance with the specific problems at hand. Missing thermodynamic data are substituted in THEREDA by well-founded estimates, thus permitting future model calculations for safety analysis to be carried out on a clearly broader basis of data. The data are managed centrally in a database and will be available to users free of charge on the Internet. Import formats allowing THEREDA to be transferred into the most common modeling codes (EQ3/6, PHREEQC, Geochemist's Workbench, CHEMAPP, etc.) are also made available free of charge. (orig.)

Concepts and examples of safety analyses for radioactive waste repositories in continental geological formations

International Nuclear Information System (INIS)

1983-01-01

This document is addressed to authorities and specialists responsible for or involved in planning, performing and/or reviewing safety assessments of underground radioactive waste repositories. It is a companion to a general introductory document on the subject ''Safety Assessment for the Underground Disposal of Radioactive Wastes'', IAEA Safety Series No. 56, 1981, and reference to this earlier document will facilitate the reader's understanding of the present report. Since examples of safety analyses are summarized here, it is hoped that this document will contribute to providing a basis for a common understanding among authorities and specialists concerned with the numerous studies involving a variety of scientific disciplines. While providing technical information, this document is also intended to stimulate further international discussion. The purposes of this report are: a) to identify the factors to be taken into account in radiological safety analyses of deep geological repositories, indicating as far as possible their relative importance during the various phases of system development; b) to show how these factors have been analysed in various safety assessment studies; and c) to comment on the merits of the selected and alternative approaches

Concepts and examples of safety analyses for radioactive waste repositories in continental geological formations

Energy Technology Data Exchange (ETDEWEB)

1983-01-01

This document is addressed to authorities and specialists responsible for or involved in planning, performing and/or reviewing safety assessments of underground radioactive waste repositories. It is a companion to a general introductory document on the subject ''Safety Assessment for the Underground Disposal of Radioactive Wastes'', IAEA Safety Series No. 56, 1981, and reference to this earlier document will facilitate the reader's understanding of the present report. Since examples of safety analyses are summarized here, it is hoped that this document will contribute to providing a basis for a common understanding among authorities and specialists concerned with the numerous studies involving a variety of scientific disciplines. While providing technical information, this document is also intended to stimulate further international discussion. The purposes of this report are: a) to identify the factors to be taken into account in radiological safety analyses of deep geological repositories, indicating as far as possible their relative importance during the various phases of system development; b) to show how these factors have been analysed in various safety assessment studies; and c) to comment on the merits of the selected and alternative approaches.

Safety analysis of a sub-seabed repository of HLW

International Nuclear Information System (INIS)

Karpf, A.D.

1989-01-01

This national safety study of a repository in the Atlantic Ocean uses as much input from German research work as possible. It is based on a computer code IMPONADOR which can be run on a personal computer. It also introduces a new way of presenting information about radionuclide confinement to individual compartments to facilitate the understanding of release processes. The work is part of the international cooperation within the Seabed Working Group of the OECD/NEA

Project Alternative Systems Study - PASS. Analysis of performance and long-term safety of repository concepts

International Nuclear Information System (INIS)

Birgersson, L.; Skagius, K.; Wiborgh, M.; Widen, H.

1992-09-01

This study is part of the Project on Alternative Systems Study, PASS, with the overall aim to perform a technical/economical ranking of alternative repository concepts and canisters for the final storage of spent nuclear fuel. The comparison should in the first stage separately assess technology in construction and operation, long-term performance and safety, and costs. Three of the repository concepts are assumed to be located at a depth of approximately 500 m in the host rock, KBS-3, Very Long Holes (VLH) and Medium Long Holes (MLH). In the KBS-3 concept the canisters are deposited in vertical deposition holes in a system of parallel storage tunnels. In the VLH concept larger canisters are deposited in long horizontal tunnels. The MLH concept, is an evolution of the two other concepts, with KBS-3 type canisters deposited in horizontal tunnels. Smaller canisters are to be deposited in deep bore holes at a depth between 2000 to 4000 m in the Very Deep Holes (VDH) concept. In all concepts the canisters will be surrounded by a bentonite buffer. The aim of the present study is to analyze and compare the performance and long-term safety of the repository concepts. Only a qualitative comparison of the concepts is made as no calculations of radionuclide releases or dose to man have been performed. The ranking of the repository concepts was carried out by comparing the VDH, VLH and MLH concept with the KBS-3 concept. The performance and long-term safety of the repositories located at 500 m level will be based on a multiple barrier system and the predictions for the concepts will involve similar uncertainties. (54 refs.)

Workshop on Regulatory Review and Safety Assessment Issues in Repository Licensing

Energy Technology Data Exchange (ETDEWEB)

Wilmot, Roger D. (Galson Sciences Limited (United Kingdom))

2011-02-15

The workshop described here was organised to address more general issues regarding regulatory review of SKB's safety assessment and overall review strategy. The objectives of the workshop were: - to learn from other programmes' experiences on planning and review of a license application for a nuclear waste repository, - to offer newly employed SSM staff an opportunity to learn more about selected safety assessment issues, and - to identify and document recommendations and ideas for SSM's further planning of the licensing review

NAGRA - Sites for geological repositories - Technical safety factors: Suggestions for stage 3

International Nuclear Information System (INIS)

2015-01-01

This comprehensive brochure published by the Swiss National Cooperative for the Disposal of Radioactive Waste (NAGRA) examines the six sites for repositories for nuclear wastes in Switzerland which have been proposed in Stage 1 of the program concerning nuclear waste repositories. Three of these sites are proposed for both highly radioactive wastes as well as for low and medium-active wastes, the other three for low and medium-active wastes only. The evaluation of the sites is discussed. The sites are to be further evaluated in Stage 2 of the program. The work to be done in the further stages involved in the selection of the final site (or sites) is described. Along with definition of the regions where deep repositories could possibly be built, suggestions for the placing of the facilities required on the surface are discussed. Geological requirements on the repositories and safety-relevant characteristics of the various site options are discussed. The results of the assessments made are presented in tabular form. Maps and geological cross-sections of all the suggested areas are included

Aerosol particle transport modeling for preclosure safety studies of nuclear waste repositories

International Nuclear Information System (INIS)

Gelbard, F.

1989-01-01

An important concern for preclosure safety analysis of a nuclear waste repository is the potential release to the environment of respirable aerosol particles. Such particles, less than 10 μm in aerodynamic diameter, may have significant adverse health effects if inhaled. To assess the potential health effects of these particles, it is not sufficient to determine the mass fraction of respirable aerosol. The chemical composition of the particles is also of importance since different radionuclides may pose vastly different health hazards. Thus, models are needed to determine under normal and accident conditions the particle size and the chemical composition distributions of aerosol particles as a function of time and of position in the repository. In this work a multicomponent sectional aerosol model is used to determine the aerosol particle size and composition distributions in the repository. A range of aerosol mass releases with varying mean particle sizes and chemical compositions is used to demonstrate the sensitivities and uncertainties of the model. Decontamination factors for some locations in the repository are presented. 8 refs., 1 tab

Project JADE. Long-term function and safety. Comparison of repository systems

International Nuclear Information System (INIS)

Birgersson, Lars; Pers, K.; Wiborgh, M.

2001-12-01

A comparison of the KBS-3 V(ertical deposition), KBS-3 H(orizontal deposition) and MLH repository systems with regard to the long-term repository performance and the radionuclide migration is presented in the report. Several differences between the repository systems have been identified. The differences are mainly related to the: distance between canister and backfilled tunnels, excavated rock volumes, deposition hole direction. The overall conclusion is that the differences are in general quite small with regard to the repository function and safety. None of the differences are of such importance for the long-term repository performance and radionuclide migration that they discriminate any of the repository systems. The differences between the two KBS-3 systems are small. Based on this study, there is no reason to change from the reference system KBS-3 V to KBS-3 H. MLH has the potential to be a very robust system, especially in a long-term perspective. However, the MLH system will require extensive research, development, and analysis before it will be as confident as the reference repository system, KBS-3 V. Although the MLH and KBS-3 H systems are in some ways favourable compared to the reference system KBS-3 V, the overall conclusion is that the KBS-3 V system is still a very attractive system. A major advantage with KBS-3 V is that it is by far the most investigated and developed system. The JADE-project was initiated in 1996, and the main part of the study was carried out during 1997 and 1998. The JADE study is consequently based on presumptions that were valid a few years ago. Some of these presumptions have been modified since then. The new presumptions are however not judged to change the overall conclusions

Evaluation of health and safety impacts of defense high-level waste in geologic repositories

International Nuclear Information System (INIS)

Smith, E.D.; Kocher, D.C.; Witherspoon, J.P.

1985-02-01

Pursuant to the requirement of the Nuclear Waste Policy Act of 1982 that the President evaluate the use of commercial high-level waste repositories for the disposal of defense high-level wastes, a comparative assessment has been performed of the potential health and safety impacts of disposal of defense wastes in commercial or defense-only repositories. Simplified models were used to make quantitative estimates of both long- and short-term health and safety impacts of several options for defense high-level waste disposal. The results indicate that potential health and safety impacts are not likely to vary significantly among the different disposal options for defense wastes. Estimated long-term health and safety impacts from all defense-waste disposal options are somewhat less than those from commercial waste disposal, while short-term health and safety impacts appear to be insensitive to the differences between defense and commercial wastes. In all cases, potential health and safety impacts are small because of the need to meet stringent standards promulgated by the US Environmental Protection Agency and the US Nuclear Regulatory Commission. We conclude that health and safety impacts should not be a significant factor in the choice of a disposal option for defense high-level wastes. 20 references, 14 tables

Source-book of International Activities Related to the Development of Safety Cases for Deep Geological Repositories

International Nuclear Information System (INIS)

2017-01-01

All national radioactive waste management authorities recognise today that a robust safety case is essential in developing disposal facilities for radioactive waste. To improve the robustness of the safety case for the development of a deep geological repository, a wide variety of activities have been carried out by national programs and international organisations over the past years. The Nuclear Energy Agency, since first introducing the modern concept of the 'safety case', has continued to monitor major developments in safety case activities at the international level. This Source-book summarises the activities being undertaken by the Nuclear Energy Agency, the European Commission and the International Atomic Energy Agency concerning the safety case for the operational and post-closure phases of geological repositories for radioactive waste that ranges from low-level to high-level waste and for spent fuel. In doing so, it highlights important differences in focus among the three organisations

A GoldSim modeling approach to safety assessment of an LILW repository system

International Nuclear Information System (INIS)

Lee, Youn Myoung; Jeong, Jongtae; Choi, Jongwon

2011-01-01

A program for the safety assessment and performance evaluation of a low- and intermediate level waste (LILW) repository system has been developed by utilizing GoldSim. By utilizing this nuclide transport in the near- and far-field of a repository as well as a transport through a biosphere under various natural and manmade disruptive events affecting a nuclide release are modeled and evaluated. To demonstrate its usability, some illustrative cases under the selected scenarios including the influence of degradation of manmade barriers, pumping well drilling, and the natural disruptive events such as a sudden formation of preferential flow pathway have been investigated and illustrated for a hypothetical LILW repository. Even though all the parameter values applied to a hypothetical repository are assumed without any real base, the illustrative cases could be informative especially when seeing the result of the probabilistic calculation or sensitivity studies with various scenarios that possibly happen for nuclide release and further transport. (author)

Probabilistic safety assessment for a generic deep geological repository for high-level waste and long-lived intermediate-level waste in clay

International Nuclear Information System (INIS)

Resele, G.; Holocher, J.; Mayer, G.; Hubschwerlen, N.; Niemeyer, M.; Beushausen, M.; Wollrath, J.

2010-01-01

Document available in extended abstract form only. In the selection procedure for the search of a final site location for the disposal of radioactive wastes, the comparison and evaluation of different potentially suitable repository systems in different types of host rocks will be an essential and crucial step. Since internationally accepted guidelines on how to perform such quantitative comparisons between repository systems with regard to their long-term safety behaviour are still lacking, in 2007 the German Federal Office for Radiation Protection launched the project 'VerSi' (Vergleichende Sicherheitsanalysen - Comparing Safety Assessments) that aims at the development of a methodology for the comparison of long-term safety assessments. A vital part of the VerSi project is the performance of long-term safety assessments for the comparison of two repository systems. The comparison focuses on a future repository for heat-generating, i.e. high-level and long-lived intermediate-level radioactive wastes in Germany. Rock salt is considered as a potential host rock for such a repository, and one repository system in VerSi is defined similarly to the potential site located in the Gorleben salt dome. Another suitable host rock formation may be clay. A generic location within the lower Cretaceous clays in Northern Germany is therefore chosen for the comparison of safety assessments within the VerSi project. The long-term safety assessment of a repository system for heat-generating radioactive waste at the generic clay location comprises different steps, amongst others: - Identifying the relevant processes in the near-field, in the geosphere and in the biosphere which are relevant for the long-term safety behaviour. - Development of a safety concept for the repository system. - Deduction of scenarios of the long-term evolution of the repository system. - Definition of statistic weights, i. e. the likelihood of occurrence of the scenarios. - Performance of a

Identification of structures, systems, and components important to safety at the potential repository at Yucca Mountain

International Nuclear Information System (INIS)

Hartman, D.J.; Miller, D.D.; Klamerus, L.J.

1991-10-01

This study recommends which structures, systems, and components of the potential repository at Yucca Mountain are important to safety. The assessment was completed in April 1990 and uses the reference repository configuration in the Site Characterization Plan Conceptual Design Report and follows the methodology required at that time by DOE Procedure AP6.10-Q. Failures of repository items during the preclosure period are evaluated to determine the potential offsite radiation doses and associated probabilities. Items are important to safety if, in the event they fail to perform their intended function, an accident could result which causes a dose commitment greater than 0.5 rem to the whole body or any organ of an individual in an unrestricted area. This study recommends that these repository items include the structures that house spent fuel and high-level waste, the associated filtered ventilation exhaust systems, certain waste- handling equipment, the waste containers, the waste treatment building structure, the underground waste transporters, and other items listed in this report. This work was completed April 1990. 27 refs., 7 figs., 9 tabs

Safety case for license application for a final repository: The French example

International Nuclear Information System (INIS)

Boissier, Fabrice; Voinis, Sylvie

2014-01-01

The reversible repository in a deep geological formation is the French reference solution for the long-term management of high-level and intermediate-level long-lived radioactive waste (HLW and ILW). Twenty years of R and D work and conceptual and basic studies since the first French Act of 1991 led, in particular, to a feasibility demonstration in 2005. According to the French Act on Radioactive Waste of 28 of June 2006, Andra shall design a reversible repository in order to apply for license in 2015. In response to this demand, Andra developed the industrial project known as 'Cigeo', a reversible geological disposal facility for HLW and ILW located in Meuse/Haute-Marne. Two years before applying for authorisation, Andra's project is now focusing on three main targets: developing Cigeo's industrial design, preparing the authorisation process through increased exchanges with stakeholders and the preparation of a safety case to support authorisation application. The latter draws on the previous safety cases of 2005 and 2009, which give a sound basis to assess Cigeo's safety, both for the operational and post-closure periods. In this new stage of the project, the challenging issues for the preparation of the safety case are the following: - to identify the various regulatory frameworks (nuclear and non-nuclear) and guides applicable to the facility; - to ensure that the industrial design complies in particular with the safety requirements as presented in the safety case and its supporting safety assessment; - to identify crucial inputs (R and D, tests,...) needed to support the authorisation application, in particular, to bring convincing arguments to assess the technical feasibility of the design and when appropriate its ability to meet the safety requirements; - to ensure that all the requirements from previous regulatory and peer reviews (national and international?) are taken into account. (authors)

Long-Term Safety Analysis of Baldone Radioactive Waste Repository and Updating of Waste Acceptance Criteria

International Nuclear Information System (INIS)

2001-12-01

The main objective of the project was to provide advice to the Latvian authorities on the safety enhancements and waste acceptance criteria for near surface radioactive waste disposal facilities of the Baldone repository. The project included the following main activities: Analysis of the current status of the management of radioactive waste in Latvia in general and, at the Baldone repository in particular Development of the short and long-term safety analysis of the Baldone repository, including: the planned increasing of capacity for disposal and long term storage, the radiological analysis for the post-closure period Development of the Environment Impact Statement, for the new foreseen installations, considering the non radiological components Proposal of recommendations for future updating of radioactive waste acceptance criteria Proposal of recommendations for safety upgrades to the facility. The work programme has been developed in phases and main tasks as follows. Phase 0: Project inception, Phase 1: Establishment of current status, plans and practices (Legislation, regulation and standards, Radioactive waste management, Waste acceptance criteria), Phase 2: Development of future strategies for long-term safety management and recommendations for safety enhancements. The project team found the general approach use at the installation, the basic design and the operating practices appropriate to international standards. Nevertheless, a number of items subject to potential improvements were also identified. These upgrading recommendations deal with general aspects of the management (mainly storage versus disposal of long-lived sources), site and environmental surveillance, packaging (qualification of containers, waste characterization requirements), the design of an engineered cap and strategies for capping. (author)

Consideration on safety assessment methodologies applied to the near surface repository Baita Bihor

International Nuclear Information System (INIS)

Dogaru, D.

2003-01-01

The Romanian legislation in respect of RAW management is described. The waste facilities in the country are: for low and intermediate level waste - Radioactive Waste Treatment Plant - Bucharest Magurele; Radioactive Waste Treatment Plant - Pitesti; National Repository for Radioactive Waste - Baita Bihor. for spent fuel - Intermediate dry spent fuel storage facility (DICA) - CNE Cernavoda; Intermediate wet spent fuel storage facility WWR-S - Bucharest Magurele. A detailed description of the facilities and waste characterisation are given in the report. Due o insufficient and incomplete information about site characterisation and inventory a Phare project 'Preliminary Safety Analysis for the Low-Level Radioactive Waste Repository Baita Bihor, Romania' has been approved. The project purposes are: to achieve a database with specific parameters; validation of scenarios and conceptual models for normal and altered evolution of the disposal site; validation and qualification of existing calculation methods and identification of the complementary suitable computer codes to be installed in Romania; validation and analyses of the final results expertise PSAR final results; recommendation for further completion of Integrated Performance Assessment. The results, conclusions and recommendations of the project will be included in the Preliminary Safety Analyses Report to be sent to the Romanian Authority - CNCAN for licensing of the repository operation

Preliminary safety assessment study for the conceptual design of a repository in tuff at Yucca Mountain

International Nuclear Information System (INIS)

Jackson, J.L.; Gram, H.F.; Hong, K.J.; Ng, H.S.; Pendergrass, A.M.

1984-12-01

Preliminary estimates of the upper bounds on postulated worst-case radiological releases resulting from possible accidents during the operating period of a prospective repository in tuff at Yucca Mountain are presented. Possible disrupting events are screened to identify the accidents of greatest potential consequence. The radiological dose commitments for the general public and repository personnel are estimated for postulated releases caused by natural phenomena, man-made events, and operational accidents. All postulated worst-case releases result in doses to the public that are lower than the 0.5-rem, whole-body dose-per-accident limit set by the Nuclear Regulatory Commission (NRC) in 10 CFR 60. Doses to repository personnel are within the NRC's 5.0-rem/yr occupational exposure limit set in 10 CFR 20 for normal operations. Doses are within this limit for all accidents except the transportation accident and fire in a drift. A preliminary risk assessment has also been performed. Based on this preliminary safety study, the proposed site boundaries and design criteria routinely used in constructing nuclear facilities appear to be adequate to protect the safety of the general public during the operating phase of the repository

Preliminary safety evaluation of an aircraft impact on a near-surface radioactive waste repository

Energy Technology Data Exchange (ETDEWEB)

Lo Frano, R.; Forasassi, G.; Pugliese, G. [Department of Industrial and Civil Engineering (DICI), University of Pisa, Pisa (Italy)

2013-07-01

The aircraft impact accident has become very significant in the design of a nuclear facilities, particularly, after the tragic September 2001 event, that raised the public concern about the potential damaging effects that the impact of a large civilian airplane could bring in safety relevant structures. The aim of this study is therefore to preliminarily evaluate the global response and the structural effects induced by the impact of a military or commercial airplane (actually considered as a 'beyond design basis' event) into a near surface radioactive waste (RWs) disposal facility. The safety evaluation was carried out according to the International safety and design guidelines and in agreement with the stress tests requirements for the security track. To achieve the purpose, a lay out and a scheme of a possible near surface repository, like for example those of the El Cabril one, were taken into account. In order to preliminarily perform a reliable analysis of such a large-scale structure and to determine the structural effects induced by such a types of impulsive loads, a realistic, but still operable, numerical model with suitable materials characteristics was implemented by means of FEM codes. In the carried out structural analyses, the RWs repository was considered a 'robust' target, due to its thicker walls and main constitutive materials (steel and reinforced concrete). In addition to adequately represent the dynamic response of repository under crashing, relevant physical phenomena (i.e. penetration, spalling, etc.) were simulated and analysed. The preliminary assessment of the effects induced by the dynamic/impulsive loads allowed generally to verify the residual strength capability of the repository considered. The obtained preliminary results highlighted a remarkable potential to withstand the impact of military/large commercial aircraft, even in presence of ongoing concrete progressive failure (some penetration and spalling of the

Central repository for low- and intermediate-level waste (ALMA) conceptual design, siting and safety study

International Nuclear Information System (INIS)

Kjellbert, N.; Haeggblom, H.; Cederstroem, M.; Lundgren, T.

1980-07-01

A generic design, siting and safety study of a proposed repository for low- and intermediate-level waste has been made. Special emphasis has been placed on safety characterostics. The conceptual design and the generic site, on which the study is based, are realistically chosen in accordance with present construction techniques and the existing geohydrological conditions in Sweden. (Auth.)

Redox processes in the safety case of deep geological repositories of radioactive wastes. Contribution of the European RECOSY Collaborative Project

International Nuclear Information System (INIS)

Duro, L.; Bruno, J.; Grivé, M.; Montoya, V.; Kienzler, B.; Altmaier, M.; Buckau, G.

2014-01-01

Highlights: • The RECOSY project produced results relevant for the Safety Case of nuclear disposal. • We classify the safety related features where RECOSY has contributed. • Redox processes effect the retention of radionuclides in all repository subsystems. - Abstract: Redox processes influence key geochemical characteristics controlling radionuclide behaviour in the near and far field of a nuclear waste repository. A sound understanding of redox related processes is therefore of high importance for developing a Safety Case, the collection of scientific, technical, administrative and managerial arguments and evidence in support of the safety of a disposal facility. This manuscript presents the contribution of the specific research on redox processes achieved within the EURATOM Collaborative Project RECOSY (REdox phenomena COntrolling SYstems) to the Safety Case of nuclear waste disposal facilities. Main objectives of RECOSY were related to the improved understanding of redox phenomena controlling the long-term release or retention of radionuclides in nuclear waste disposal and providing tools to apply the results to Performance Assessment and the Safety Case. The research developed during the project covered aspects of the near-field and the far-field aspects of the repository, including studies relevant for the rock formations considered in Europe as suitable for hosting an underground repository for radioactive wastes. It is the intention of this paper to highlight in which way the results obtained from RECOSY can feed the scientific process understanding needed for the stepwise development of the Safety Case associated with deep geological disposal of radioactive wastes

USA NRC/RSR Data Bank System and Reactor Safety Research Data Repository (RSRDR)

International Nuclear Information System (INIS)

Maskewitz, B.F.; Bankert, S.F.

1979-01-01

The United States Nuclear Regulatory Commission (NRC), through its Division of Reactor Safety Research (RSR) of the Office of Nuclear Regulatory Research, has established the NRC/RSR Data Bank Program to collect, process, and make available data from the many domestic and foreign water reactor safety research programs. An increasing number of requests for data and/or calculations generated by NRC Contractors led to the initiation of the program which allows timely and direct access to water reactor safety data in a manner most useful to the user. The program consists of three main elements: data sources, service organizations, and a data repository

Model of evolution of radioactive waste repositories and their influence on the resource-ecological safety of an adjoining territories

International Nuclear Information System (INIS)

Angelova, R.; Sandul, G.A.; Sen'ko, T.Ya.

2002-01-01

In this paper it is considered the mathematical model of evolution of radioactive waste (RAW) repositories and their influence on the resource-ecological safety (RES) and sustainable development of an adjoining territories. Heart of considered model consists of that RAW repository is considered as a system with two processes proceeding in parallel: deterioration of repository buildings, equipment etc. enlarging resource-ecological danger (RED) on account of probability increase (risk increase) of emergency conditions; natural decay of RAW being in repository that lead to RED decrease. Considered model allows to learn RAW repositories evolution in given time interval and to analyze their behavior at its different stages depending on state of repositories, e.g., their modernization or other events as well as to define periods of RAW repositories peak danger for environment

Deep repository for spent nuclear fuel. SR 97 - Post-closure safety. Main Report. Summary

International Nuclear Information System (INIS)

Hedin, A.

1999-11-01

In preparation for coming site investigations for siting of a deep repository for spent nuclear fuel, the Swedish Government and nuclear regulatory authorities have requested an assessment of the repository's long-term safety. The purpose is to demonstrate whether the risk of harmful effects in individuals in the vicinity of the repository complies with the acceptance criterion formulated by the Swedish regulatory authorities, i.e. that the risk may not exceed 10 -6 per year. Geological data are taken from three sites in Sweden to shed light on different conditions in Swedish granitic bedrock. The future evolution of the repository system is analyzed in the form of five scenarios. The first is a base scenario where the repository is postulated to be built entirely according to specifications and where present-day conditions in the surroundings are postulated to persist. The four other scenarios show how the evolution of the repository differs from that in the base scenario if the repository contains a few initially defective canisters, in the event of climate change, earthquakes, and future inadvertent human intrusion. The time horizon for the analyses is at most one million years, in accordance with preliminary regulations. By means of model studies and calculations, the base scenario analyzes how the radioactivity of the fuel declines with time, the repository's thermal evolution as a result of the decay heat in the fuel, the hydraulic evolution in buffer and backfill when they become saturated with water, and the long-term groundwater flow in the geosphere on the three sites. The overall conclusion of the analyses in the base scenario is that the copper canisters isolating capacity is not threatened by either the mechanical or chemical stresses to which it is subjected. The safety margins are great even in a million-year perspective. The internal evolution in initially defective canisters and the possible resultant migration of radionuclides in buffer, geosphere

SR 97: Post-closure safety for a KBS 3 type deep repository for spent nuclear fuel

International Nuclear Information System (INIS)

Hedin, A.; Kautsky, U.

2000-03-01

Prior to coming site investigations for siting of a deep repository for spent nuclear fuel, SKB has carried out the long-term safety assessment SR 97, requested by the Swedish Government. The repository is of the KBS-3 type, where the fuel is placed in isolating copper canisters with a high-strength cast iron insert. The canisters are surrounded by bentonite clay in individual deposition holes at a depth of 500 m in granitic bedrock. The future evolution of the repository system is analysed in the form of five scenarios. The first is a base scenario where the repository is postulated to be built entirely according to specifications and where present-day conditions in the surroundings, including climate, persist. The four other scenarios show the evolution if the repository contains a few initially defective canisters, in the event of climate change, in the event of earthquakes, and in the event of future inadvertent human intrusion. The principal conclusion of the assessment is that the prospects of building a safe deep repository for spent nuclear fuel in Swedish granitic bedrock are very good. (author)

Application of geostatistical methods to long-term safety analyses for radioactive waste repositories

International Nuclear Information System (INIS)

Roehlig, K.J.

2001-01-01

Long-term safety analyses are an important part of the design and optimisation process as well as of the licensing procedure for final repositories for radioactive waste in deep geological formations. For selected scenarios describing possible evolutions of the repository system in the post-closure phase, quantitative consequence analyses are performed. Due to the complexity of the phenomena of concern and the large timeframes under consideration, several types of uncertainties have to be taken into account. The modelling work for the far-field (geosphere) surrounding or overlaying the repository is based on model calculations concerning the groundwater movement and the resulting migration of radionuclides which possibly will be released from the repository. In contrast to engineered systems, the geosphere shows a strong spatial variability of facies, materials and material properties. The paper presented here describes the first steps towards a quantitative approach for an uncertainty assessment taking into account this variability. Due to the availability of a large amount of data and information of several types, the Gorleben site (Germany) has been used for a case study in order to demonstrate the method. (orig.)

Morsleben repository - Interdependence of technical feasibility and functionality of geotechnical barriers and safety case development

International Nuclear Information System (INIS)

Wollrath, Juergen; Mauke, Ralf; Mohlfeld, Matthias; Niemeyer, Matthias; Becker, Dirk-Alexander

2014-01-01

Based on a selection procedure whereby ten existing mines had been taken into consideration, the Morsleben repository for radioactive waste (ERAM) was built in a former mine for potash and rock salt production. The specific concerns and objectives of a repository for radioactive waste could not be taken into account when the mine was built at the beginning of the last century. Irrespective of this, altogether about 37 000 m 3 of low-level and intermediate-level radioactive waste was stored in several areas of the mine between 1971 and 1991 and from 1994 to 1998. In the scope of the ongoing licensing procedure, the safety of the 'historically grown' repository needs to be demonstrated for the phase after it has been sealed. (authors)

Safety evaluation methodology of engineering barriers at repository for low and intermediate level radioactive waste

International Nuclear Information System (INIS)

Zarnic, R.; Bokan Bosiljkov, V.; Giacomelli, M.

2007-01-01

Analyses of the roles of cement-based barriers in radioactive waste isolation show that models used to estimate their characteristics during the lifetime of the repository must consider the alteration of material properties with time due to degradation processes. Reinforced concrete barriers at repositories shall be designed in such manner that they fulfil besides isolative capabilities also the required functions of mechanical resistance and stability. Key elements of safety evaluation are mainly the correct selection of materials for mineral composites with cement binder (cements, aggregates, mineral additives and chemical admixtures) and their design, execution of construction works and production of precast concrete containers (continuous casting of concrete - no cold joints, limited number of construction joints, proper placing and consolidation, finishing and curing), strict control of used materials and inspection of works, as well as investigation after the construction (visual inspection, non-destructive testing, monitoring, ageing assessment on test containers). According to the methodology presented in this paper the lifetime of the repository can be estimated and, if shorter than 300 years or shorter than the period resulting from safety analysis, appropriate corrective measures shall be taken. (author)

Communication on the Safety Case for a Deep Geological Repository

International Nuclear Information System (INIS)

Bailey, Lucy; Bernier, Frederik; Bollingerfehr, Wilhelm; Cunado, Miguel; Ilett, Doug; Kwong, Gloria; ); Noseck, Ulrich; Roehlig, Klaus; Van Luik, Abe; Weber, Jan; Weetjens, Eef

2017-01-01

Communication has a specific role to play in the development of deep geological repositories. Building trust with the stakeholders involved in this process, particularly within the local community, is key for effective communication between the authorities and the public. There are also clear benefits to having technical experts hone their communication skills and having communication experts integrated into the development process. This report has compiled lessons from both failures and successes in communicating technical information to non-technical audiences. It addresses two key questions in particular: what is the experience base concerning the effectiveness or non-effectiveness of different tools for communicating safety case results to a non-technical audience and how can communication based on this experience be improved and included into a safety case development effort from the beginning? (authors)

Methodology and applicability of a safety and demonstration concept for a HAW final repository on clays. Safety concept and verification strategy

International Nuclear Information System (INIS)

Ruebel, Andre; Meleshyn, Artur

2014-08-01

The report describes the site independent frame for a safety concept and verification strategy for a final repository for heat generating wastes in clay rock. In the safety concept planning specifications and technical measures are summarized that are supposed to allow a safe inclusion of radionuclides in the host rock. The verification strategy defines the systematic procedures for the development of fundamentals and scenarios as basis for the demonstration of the safety case and to allow the prognosis of appropriateness. The report includes the boundary conditions, the safety concept for the post-closure phase and the verification strategy for the post-closure phase.

Safety Assessment for a KBS-3H spent nuclear fuel repository at Olkiluoto

International Nuclear Information System (INIS)

Smith, P.; Neall, F.; Snellman, M.; Pastina, B.; Hjerpe, T.; Nordman, H.; Johnson, L.

2007-12-01

The KBS-3 method, based on multiple barriers, is the proposed spent fuel disposal method both in Sweden and Finland. KBS-3H and KBS-3V are the two design alternatives of the KBS-3 spent fuel disposal method. Posiva and SKB have conducted a joint research, demonstration and development (RD and D) programme in 2002-2007 with the overall aim of establishing whether KBS-3H represents a feasible alternative to the reference alternative KBS-3V. The overall objectives of the present phase covering the period 2004-2007 have been to demonstrate that the horizontal deposition alternative is technically feasible and to demonstrate that it fulfils the same long-term safety requirements as KBS-3V. The safety studies conducted as part of this programme include a safety assessment of a preliminary design of a KBS-3H repository for spent nuclear fuel located about 400 m underground at the Olkiluoto site, which is the proposed site for a spent fuel repository in Finland. This safety assessment is summarised in the present report. The scientific basis of the safety assessment includes around 30 years of scientific R and D and technical development in the Swedish and Finnish KBS-3V programmes. Much of this scientific basis is directly applicable to KBS-3H. This has allowed the KBS-3H safety studies to focus on those issues that are unique to this design alternative, identified in a systematic difference analysis of KBS-3H and KBS-3V. This difference analysis has shown that the key differences in the evolution and performance of KBS-3H and KBS-3V relate mainly to the engineered barrier system and to the impact of local variations in the rate of groundwater inflow on buffer saturation along the KBS-3H deposition drifts. No features or processes specific to KBS-3H have been identified that could lead to a loss or substantial degradation of the safety functions of the engineered barriers over a million year time frame. Radionuclide release from the repository near field in the event of

Fuzzy Logic Inference System for Determining The Quality Assesment of Student’s Learning ICT

Directory of Open Access Journals (Sweden)

Agus Pamuji

2017-05-01

Full Text Available The Assesment that held in the school is one of the learning process in education who do it by teacher. One of the course that exemined is Computer Application. In the computer application have 3 topic, they are Microsoft Word, Microsoft Excel, Microsoft Power Point. The assesment for student’s at politecnic about learning computer application have 3 criteria in the selection. First of all, the students have ability to operate computer system generaly, it has understanding the formula on microsoft excel, the students have skill toward any application. In this study, fuzzy logic used for determining the quality assesment of stundent’s learning Information and Comunication Technology (ICT as a tools to analyze any constraint that are known as min-max method. As a result, we have found that the students have good for analyzing in the application from the each question or case of study when the course it has been examined. 

Current trends in degradation assesment on metallic materials of industrial components

International Nuclear Information System (INIS)

Herrera Palma, Victoria

2007-01-01

To needs to assess objectively a structural integrity analysis in nuclear and termal power-, oil- and chemical- industry system, represents a large challenge for engineer and researches related to Materials Science, equipment manufactures or users. These systems share many of their problems with regards to aging mechanism of components metallic materials, high replacement costs and increasing requirements on efficiency and safety. This paper makes an attempt to give an overview of the current trends on material damage and residual life assessment for installation of power-, oil- and chemical industry. Some of the currently existing ideas on components inspection, as an activity for damage detection are shown. A summary on mechanism of material damage and experimental techniques for their characterization is also presented. Finally, some analytical methods with wide appliance in materials damage evaluation and residual life assesment of components are described

Making the post-closure safety case for the proposed yucca mountain repository

International Nuclear Information System (INIS)

Swift, P.; Van Luik, A.

2008-01-01

This presentation provided an overview of the Yucca Mountain repository post-closure safety case. The safety case concept is being integrated into the license application being prepared for Yucca Mountain, by giving particularly close attention to the treatment of uncertainties, thereby bringing available lines of evidence into the supporting information, as appropriate, to build a comprehensive argument for safety and regulatory compliance. For Yucca Mountain, it is expected that there will be open questions in the safety case to be presented to the regulator and a programme will be outlined on what information is to be gathered (and how) prior to the next iteration in the licensing process to address such open issues. A one-hundred year operational phase is foreseen and planned, and the changes in knowledge and approaches that occur over time will have to be accommodated through the formal licensing process. (authors)

Safety Report within the licence application for the siting of a radioactive waste repository/disposal facility

International Nuclear Information System (INIS)

Horyna, J.; Sinaglova, R.

2004-01-01

The initial safety specification report, which is submitted to the licensing authority as one of the application documents, is the basic document assessing the planned repository/disposal facility with respect to the suitability of the chosen site for this purpose. The following topics are covered: General information; Description and evidence of suitability of the site chosen; Description and tentative assessment of the repository/disposal facility design; Tentative assessment of impacts of running the facility on the employees, general public and environment (radionuclide inventory, transport routes, radionuclide release in normal, abnormal and emergency situations); Proposed concept of repository/disposal facility shutdown; and Assessment of quality assurance in the site selection, in preparatory work for the construction of the facility and in the subsequent stages. (P.A.)

The development of safeguards for geological repositories

International Nuclear Information System (INIS)

Van der Meer, K.

2009-01-01

Traditionally, research and development on geological repositories for High Level Waste (HLW) focuses on the short- and long-term safety aspects of the repository. If the repository will also be used for the disposal of spent fuel, safeguards aspects have to be taken into account. Safety and safeguards requirements may be contradictory; the safety of a geological repository is based on the non-intrusion of the geological containment, while safeguards require regular inspections of position and amount of the spent fuel. Examples to reconcile these contradictory requirements are the use of information required for the safety assessment of the geological repository for safeguards purposes and the adaptation of the safeguards approach to use non-intrusive inspection techniques. The principles of an inspection approach for a geological repository are now generally accepted within the IAEA. The practical applicability of the envisaged inspection techniques is still subject to investigation. It is specifically important for the Belgian situation that an inspection technique can be used in clay, the geological medium in which Belgium intends to dispose its HLW and spent fuel. The work reported in this chapter is the result of an international cooperation in the framework of the IAEA, in which SCK-CEN participates

Recommendations: Procedure to develop a preliminary safety report as part of the radioactive waste repository construction licensing process

International Nuclear Information System (INIS)

2003-01-01

The structure of a preliminary safety report for the title purpose should be as follows: A. Textual part: 1. General (Introduction, Basic information about the construction, Timetable); 2. Site information (Siting, Geography and demography, Meteorology and climatic situation, Hydrology, Geology and hydrogeology); 3. Repository design description (Basic function and performance requirements, Design, Auxiliary systems, Fire prevention/protection, Emergency plans); 4. Operation of the repository (Waste acceptance and inspection, Waste handling and interim storage, Waste disposal, Operating monitoring), 5. Health and environmental impact assessment (Radionuclide inventory, Radionuclide transport paths and mechanisms of release into the environment, Radionuclide release in normal and emergency situations, Radiation protection - health impact assessment and regulatory compliance, Draft operating limits and conditions, Proposed ways of assuring physical protection, Uncertainty assessment), 6. Safe repository shutdown/decommissioning concept, 7 Quality assurance assessment, 8. List of selected equipment. B. Annexes: Maps, Drawings, Diagrams, Miscellaneous; C. Documentation: Previous safety report amendments, Protocols, Miscellaneous. (P.A.)

The safety case in support of the license application of the surface repository of low-level waste in Dessel, Belgium

International Nuclear Information System (INIS)

Wacquier, William; Cool, Wim

2014-01-01

The modern concept of the safety case, developed by the OECD/NEA for geological repositories of high- and medium-level waste has been successfully applied by ONDRAF/ NIRAS for a surface repository for Category A waste (i.e. low-level waste) in Belgium in the current project phase 2006-2012. This resulted in the submission on 31 January 2013 by ONDRAF/NIRAS of an application for a 'construction and operation license' to the safety authorities. The benefits of using the notion of the safety case have been that: i) safety has been incorporated in an integrated manner within all assessment basis, design and safety assessment activities; ii) the process of development of the license application has gained in clarity and traceability; iii) the documentation of the license application contains multiple lines of argumentation for safety rather than argumentation based only on quantitative radiological impact calculations. To offer a comprehensive view on the safety argumentation and its development, it has been found useful to develop the argumentation not only along a safety statements structure but also along the safety report structure. (authors)

Safety and performance indicators for repositories in salt and clay formations

International Nuclear Information System (INIS)

Wolf, Jens; Ruebel, Andre; Noseck, Ulrich; Becker, Dirk

2008-07-01

The GRS (Gesellschaft fuer Reaktorsicherheit) study aims to the identification of suitable indicators for repositories in salt and clay formation. It is not intended to compare the two formations with respect to the safe disposal of radioactive waste. A first set of safety and performance indicators for both host rocks has been derived on the basis of results of the SPIN project. Reference values for the safety indicators have been determined. The suitability of the indicators and their significance for different time frames Is demonstrated by means of deterministic model calculations and external parameter variations of previous studies. The safety indicators considered in the report are the effective dose rate (Sv/a), the radiotoxicity concentration in the biosphere water (Sv/m 3 ) and the radiotoxicity flux from the geosphere (overlying rock) (Sv/a). The performance indicators considered in the study are the radiotoxicity inventory in different compartments (S), radiotoxicity fluxes from compartments and the integrated radiotoxicity fluxes from compartments (Sv).

Safety- and performance indicators for a generic deep geological repository in clay

International Nuclear Information System (INIS)

Resele, G.; Niemeyer, M.; Wilhelm, St.; Heimer, St.; Mohlfeld, M.; Eilers, G.; Preuss, J.; Wollrath, J.

2010-01-01

Document available in extended abstract form only. As a first step of an impartial survey for an optimal site selection for a deep geological repository in Germany, potentially suitable regions shall be identified and localised according to their suitability. During the early phases of such a site selection procedure the information about the properties of the host rock and the geological situation at the potential sites is not very precise. As site investigation procedures are both expensive and time-consuming, it is essential to identify those properties of the geological barrier system that are most relevant for long-term safety. Furthermore, adequate indicators have to be chosen that allow a simple but efficient assessment of the suitability of the potential regions. Definition and application of 'exclusion criteria' based on single parameter values, e.g. the hydraulic conductivity of the host rock, is inadequate because the long-term safety depends on the interaction of many features and properties of the barrier system. In a research project, indicators have been developed which depend on the most relevant properties of the geological barriers and estimate the overall performance of a repository system. The application of these indicators on the barrier properties which have been found during the investigations of potential repository sites in clay located in Germany, Switzerland and France demonstrates how, for instance, an unfavourably high hydraulic permeability of the clay can be compensated by a large vertical extension of the clay layer and small hydraulic gradients. Other indicators evaluate the importance of hydraulic discontinuities and define the minimal requirements on technical barriers like seals and backfill of emplacement tunnels. When the information of the radionuclide inventory and the biosphere, especially the diluting aquifer is included, the indicators allow the estimation of the resulting dose which matches the result of a

An innovative 3-D numerical modelling procedure for simulating repository-scale excavations in rock - SAFETI

Energy Technology Data Exchange (ETDEWEB)

Young, R. P.; Collins, D.; Hazzard, J.; Heath, A. [Department of Earth Sciences, Liverpool University, 4 Brownlow street, UK-0 L69 3GP Liverpool (United Kingdom); Pettitt, W.; Baker, C. [Applied Seismology Consultants LTD, 10 Belmont, Shropshire, UK-S41 ITE Shrewsbury (United Kingdom); Billaux, D.; Cundall, P.; Potyondy, D.; Dedecker, F. [Itasca Consultants S.A., Centre Scientifique A. Moiroux, 64, chemin des Mouilles, F69130 Ecully (France); Svemar, C. [Svensk Karnbranslemantering AB, SKB, Aspo Hard Rock Laboratory, PL 300, S-57295 Figeholm (Sweden); Lebon, P. [ANDRA, Parc de la Croix Blanche, 7, rue Jean Monnet, F-92298 Chatenay-Malabry (France)

2004-07-01

This paper presents current results from work performed within the European Commission project SAFETI. The main objective of SAFETI is to develop and test an innovative 3D numerical modelling procedure that will enable the 3-D simulation of nuclear waste repositories in rock. The modelling code is called AC/DC (Adaptive Continuum/ Dis-Continuum) and is partially based on Itasca Consulting Group's Particle Flow Code (PFC). Results are presented from the laboratory validation study where algorithms and procedures have been developed and tested to allow accurate 'Models for Rock' to be produced. Preliminary results are also presented on the use of AC/DC with parallel processors and adaptive logic. During the final year of the project a detailed model of the Prototype Repository Experiment at SKB's Hard Rock Laboratory will be produced using up to 128 processors on the parallel super computing facility at Liverpool University. (authors)

The waste isolation pilot plant transuranic waste repository: A case study in radioactive waste disposal safety and risk

Energy Technology Data Exchange (ETDEWEB)

Eriksson, Leif G. [GRAM, Inc., Albuquerque, NM (United States)

1999-12-01

The Waste Isolation Pilot Plant (WIPP) deep geological defense-generated transuranic radioactive waste (TRUW) repository in the United States was certified on the 13 of May 1998 and opened on the 26 of March 1999. Two sets of safety/performance assessment calculations supporting the certification of the WIPP TRUW repository show that the maximum annual individual committed effective dose will be 32 times lower than the regulatory limit and that the cumulative amount of radionuclide releases will be at least 10 times, more likely at least 20 times, lower than the regulatory limits. Yet, perceptions remain among the public that the WIPP TRUW repository imposes an unacceptable risk.

The waste isolation pilot plant transuranic waste repository: A case study in radioactive waste disposal safety and risk

International Nuclear Information System (INIS)

Eriksson, Leif G.

1999-01-01

The Waste Isolation Pilot Plant (WIPP) deep geological defense-generated transuranic radioactive waste (TRUW) repository in the United States was certified on the 13 of May 1998 and opened on the 26 of March 1999. Two sets of safety/performance assessment calculations supporting the certification of the WIPP TRUW repository show that the maximum annual individual committed effective dose will be 32 times lower than the regulatory limit and that the cumulative amount of radionuclide releases will be at least 10 times, more likely at least 20 times, lower than the regulatory limits. Yet, perceptions remain among the public that the WIPP TRUW repository imposes an unacceptable risk

Safety case for Slovenian LILW near-surface repository

International Nuclear Information System (INIS)

Viršek, Sandi; Špiler, Janja; Žagar, Tomaž

2016-01-01

Conclusions: • Repository provides the fulfillment of international and national legal requirements regarding treatment and disposal of LILW; • Repository improves the conditions for life extension for NPP Krško and offers synergetic effects for the second unit of NPP Krško; • Repository provides a basis for safe, economic and reliable use of radioactive sources in science, medicine and industry in Slovenia; • All economic calculations and comparisons show a clear advantage in case of a joint Slovenian and Croatian solution

Transfer systems in an underground repository

International Nuclear Information System (INIS)

Berg, H.P.; Ehrlich, D.

1991-01-01

In addition to logistic problem definitions taking into account the waste types of the wastes to be disposed of and the mining conditions, transport and handling of radioactive wastes in a repository, particularly require the keeping of safety technological marginal conditions mainly resulting from the accident analyses carried out. The realization of these safety technological aspects is described taking the planned Konrad repository as an example. (author)

Initial Q-list for the prospective Yucca Mountain repository based on items important to safety and waste isolation

International Nuclear Information System (INIS)

Laub, T.W.; Jardine, L.J.

1987-01-01

A method for identifying items important to safety based on a probabilistic risk assessment approach was developed and implemented for the conceptual design of the Yucca Mountain repository. No items were classified as important to safety; however, six items were classified as potentially important to safety. These were the shipping cask, the cranes and the truck or rail-care vehicle stops in the cask receiving and preparation area, the hot cell structure of the waste packaging hot cells, the cranes in the waste packaging hot cells, and the waste-handling building fire protection system. In addition, a method for identifying items important to waste isolation was developed and implemented. Two hydrogeologic units of the Yucca Mountain site were classified as important to waste isolation: the Calico Hills nonwelded zeolitic unit and the Calico Hills nonwelded vitric unit. The preliminary Q-list for the Yucca Mountain repository is comprised of the two units of the site classified as important to waste isolation and contains no items important to safety

Initial Q-list for the prospective Yucca Mountain repository based on items important to safety and waste isolation

International Nuclear Information System (INIS)

Laub, T.W.; Jardine, L.J.

1987-01-01

A method for identifying items important to safety based on a probabilistic risk assessment approach was developed and implemented for the conceptual design of the Yucca Mountain repository. No items were classified as important to safety; however, six items were classified as potentially important to safety. These were the shipping cask, the cranes and the truck or rail-car vehicle stops in the cask receiving and preparation area, the hot cell structure of the waste packaging hot cells, the cranes in the waste packaging hot cells, and the waste-handling building fire protection system. In addition, a method for identifying items important to waste isolation was developed and implemented. Two hydrogeologic units of the Yucca Mountain site were classified as important to waste isolation: the Calico Hills nonwelded zeolitic unit and the Calico Hills nonwelded vitric unit. The preliminary Q-list for the Yucca Mountain repository is comprised of the two units of the site classified as important to waste isolation and contains no items important to safety

Design and production of the KBS-3 repository

International Nuclear Information System (INIS)

Moren, Lena

2010-12-01

The report contains the common basis for a set of Production reports, presenting how the KBS-3 repository is designed, produced and inspected. The set of reports is included in the safety report for the KBS-3 repository and repository facility. The report presents the role of the Production reports within the safety report and their common purposes and objectives. An important part of the report is to present the background and sources to the principles to be applied in the design, the functions of the KBS-3 repository and the barrier functions the engineered barriers and rock. Further, the methodology to substantiate detailed design premises for the engineered barriers, underground openings and other parts of the KBS-3 repository is presented. The report also gives an overview of the KBS-3 system and its facilities and the production lines for the spent fuel, the engineered barriers and underground openings. Finally, an introduction to quality management, safety classification and their application is given

Design and production of the KBS-3 repository

Energy Technology Data Exchange (ETDEWEB)

Moren, Lena

2010-12-15

The report contains the common basis for a set of Production reports, presenting how the KBS-3 repository is designed, produced and inspected. The set of reports is included in the safety report for the KBS-3 repository and repository facility. The report presents the role of the Production reports within the safety report and their common purposes and objectives. An important part of the report is to present the background and sources to the principles to be applied in the design, the functions of the KBS-3 repository and the barrier functions the engineered barriers and rock. Further, the methodology to substantiate detailed design premises for the engineered barriers, underground openings and other parts of the KBS-3 repository is presented. The report also gives an overview of the KBS-3 system and its facilities and the production lines for the spent fuel, the engineered barriers and underground openings. Finally, an introduction to quality management, safety classification and their application is given

Technical expertise on the safety of the proposed geological repository sites. Planning for geological deep repositories, step 1; Sicherheitstechnisches Gutachten zum Vorschlag geologischer Standortgebiete. Sachplan geologische Tiefenlager, Etappe 1

Energy Technology Data Exchange (ETDEWEB)

NONE

2010-01-15

On October 17, 2010, on request of those Swiss government institutions responsible for the disposal of radioactive wastes, the National Co-operative for the Disposal of Radioactive Waste (NAGRA) presented its project concerning geological sites for the foreseen disposal of radioactive wastes to the Federal Authorities. According to the present disposal concept, two types of repository are foreseen: one for highly radioactive wastes (HAA) and the other for low radioactive and intermediate-level radioactive wastes (SMA). If a site fulfils the necessary conditions for both HAA as well as for SMA, a combined site for both types of waste may be chosen. As a qualified control authority in Switzerland, the Federal Nuclear Safety Inspectorate (ENSI) has to examine the quality of the NAGRA proposals from the point of view of the nuclear safety of the sites. The project for deep underground waste disposal first defines the process and the criteria according to which sites for the geological storage of all types of radioactive wastes in Switzerland have to be chosen. The choice is based on the actual knowledge of Swiss geology. After dividing the wastes into SMA and HAA, some large-scale areas are to be identified according to their suitability from the geological and tectonic points of view. NAGRA's division of waste into SMA and HAA is based on calculations of the long-term safety for a broad range of different rock types and geological situations and takes the different properties of all waste types into account. As a conclusion, a small portion of SMA has to be stored with {alpha}-toxic wastes in the HAA repository. The estimation of the total volume of wastes to be stored is based on 60 years of operation of the actual nuclear power plants, augmented with the wastes from possible replacement plants with a total power of 5 GW{sub e} during a further 60 years. The safety concept of the repository is based on passive systems using technical and natural barriers. The

Modelling approach to LILW-SL repository safety evaluation for different waste packing options

International Nuclear Information System (INIS)

Perko, Janez; Mallants, Dirk; Volckaert, Geert; Towler, George; Egan, Mike; Virsek, Sandi; Hertl, Bojan

2007-01-01

The key objective of the work described here was to support the identification of a preferred disposal concept and packaging option for low and short-lived intermediate level waste (LILW-SL). The emphasis of the assessment, conducted on behalf of the Slovenian radioactive waste management agency (ARAO), was the consideration of several waste treatment and packaging options in an attempt to identify optimised containment characteristics that would result in safe disposal, taking into account the cost-benefit of alternative safety measures. Waste streams for which alternative treatment and packaging solutions were developed and evaluated include decommissioning waste and NPP operational wastes, including drums with unconditioned ion exchange resins in over-packed tube type containers (TTCs). For decommissioning wastes, the disposal options under consideration were either direct disposal of loose pieces grouted into a vault or use of high integrity containers (HIC). In relation to operational wastes, three main options were foreseen. The first is over-packing of resin containing TTCs grouted into high integrity containers, the second option is complete treatment with hydration, neutralization, and cementation of the dry resins into drums grouted into high integrity containers and the third is direct disposal of TTCs into high integrity containers without additional treatment. The long-term safety of radioactive waste repositories is usually demonstrated with the support of a safety assessment. This normally includes modelling of radionuclide release from a multi-barrier near-surface or deep repository to the geosphere and biosphere. For the current work, performance assessment models were developed for each combination of siting option, repository design and waste packaging option. Modelling of releases from the engineered containment system (the 'near-field') was undertaken using the AMBER code. Detailed unsaturated water flow modelling was undertaken using the

Management of radioactive waste at Novi Han Repository

International Nuclear Information System (INIS)

Stefanova, I.G.; Mateeva, M.D.; Milanov, M.V.

2002-01-01

The Novi Han Repository is the only existing repository in Bulgaria for the disposal of radioactive waste from nuclear applications in industry, medicine and research. The repository was constructed in the early sixties according to the existing requirements. It was operated by the Institute for Nuclear Research and Nuclear Energy for more than thirty years without any accident or release of radioactivity to the environment, but without any investment for upgrading. As a consequence, the Bulgarian Nuclear Safety Authority temporarily stopped the operation of the repository in 1994. The measures for upgrading the Novi Han Repository, supported by the IAEA under TC Project BUL/4/005 'Increasing Safety of Novi Han Repository', are presented in this paper. They comprise: assessment of radionuclide inventory and future waste arisings, characterisation of disposal vaults, characterisation of the site, safety assessment, upgrading of the monitoring system, option study for the selection of treatment and conditioning processes and the development of a conceptual design for low and intermediate level waste processing and storage facility, immediate measures for improvement of the existing disposal vaults and infrastructure, construction of above-ground temporary storage structures, and resuming the operation of the Novi Han Repository. The necessary activities for re-licensing of the Novi Han Repository, construction of a waste processing and storage facility and a disposal facility for spent sealed sources are discussed. (author)

Nuclear waste in a repository: amount as a factor in risk duration

International Nuclear Information System (INIS)

Zen, E.

1980-01-01

The relationship between the amount of nuclear waste in a nuclear repository and the safety of the repository is examined. It is shown that the amount of a given hazardous nuclide that is potentially leachable depends on the initial amount of waste in the repository and the time that has elapsed since the repository was put into operation. Nuclear repository safety can be enhanced if repositories are designed as modular units with leach-resistant watertight compartments

Studies relating to human intrusion into a repository. Report pertaining to work package 11. Preliminary safety case of the Gorleben site (VSG)

Energy Technology Data Exchange (ETDEWEB)

Beuth, Thomas; Buhmann, Dieter; Fischer-Appelt, Klaus; Moenig, Joerg; Ruebel, Andre; Wolf, Jens [Gesellschaft fuer Anlagen- und Reaktorsicherheit mbH (GRS), Koeln (Germany); Bollingerfehr, Wilhelm; Filbert, Wolfgang [DBE Technology GmbH, Peine (Germany); Charlier, Frank [international nuclear safety engineering gmbh (nse), Aachen (Germany); Baltes, Bruno

2014-10-15

The question of the long-term safety of a repository system is inseparably linked with the intensive technical examination of the possible future evolution of the site and the repository system e. g. as a result of climatic, geologic, waste-related and repository-related processes. Here, the possible evolutions to be considered are those that have the potential to have a negative impact on the intended, furthest-possible, immediate, and lasting isolation of the radioactive waste in a defined area around the underground workings of the repository mine in salt rock, which is referred to as the containment-providing rock zone (CPRZ).

Safety assessment for a KBS-3H spent nuclear fuel repository at Olkiluoto. Evolution report

International Nuclear Information System (INIS)

Smith, P.; Johnson, L.; Snellman, M.; Pastina, B.; Gribi, P.

2007-12-01

The KBS-3 method, based on multiple barriers, is the proposed spent fuel disposal method both in Sweden and Finland. KBS-3H and KBS-3V are the two design alternatives of the KBS-3 method. Posiva and SKB have conducted a joint research, demonstration and development (RD and D) programme in 2002-2007 with the overall aim of establishing whether KBS-3H represents a feasible alternative to the reference alternative KBS-3V. The overall objectives of the present phase covering the period 2004-2007, have been to demonstrate that the horizontal deposition alternative is technically feasible and to demonstrate that it fulfils the same long-term safety requirements as KBS-3V. The safety studies conducted as part of this programme include a safety assessment of a preliminary design of a KBS-3H repository for spent nuclear fuel located about 400 m underground at the Olkiluoto site, which is the proposed site for a spent fuel repository in Finland. In the KBS-3H design alternative, each canister, with a surrounding layer of bentonite clay, is pre-packaged in a perforated steel cylinder prior to emplacement in the deposition drift; the entire assembly is called the supercontainer. Several supercontainers are positioned along parallel, 100 - 300 m long deposition drifts, which are sealed following waste emplacement using drift end plugs. Bentonite distance blocks separate the supercontainers, one from another, along the drift. Steel compartment plugs can be used to seal off drift sections with higher inflow, thus isolating the different compartments within the drift. The present report describes the repository evolution in successive time frames, including key uncertainties. The description of evolution starts with the initial conditions at the time of emplacement of the first canisters. The repository evolves through an early, transient phase to a state where evolution is far slower. Particular attention is given to describing the transient phase, since this is where most of the

Safety assessment for a KBS-3H spent nuclear fuel repository at Olkiluoto. Evolution report

International Nuclear Information System (INIS)

Smith, Paul; Johnson, Lawrence; Snellman, Margit; Pastina, Barbara; Gribi, Peter

2008-01-01

The KBS-3 method, based on multiple barriers, is the proposed spent fuel disposal method both in Sweden and Finland. KBS-3H and KBS-3V are the two design alternatives of the KBS-3 method. Posiva and SKB have conducted a joint research, demonstration and development (RDandD) programme in 2002-2007 with the overall aim of establishing whether KBS-3H represents a feasible alternative to the reference alternative KBS-3V. The overall objectives of the present phase covering the period 2004-2007, have been to demonstrate that the horizontal deposition alternative is technically feasible and to demonstrate that it fulfils the same long-term safety requirements as KBS-3V. The safety studies conducted as part of this programme include a safety assessment of a preliminary design of a KBS-3H repository for spent nuclear fuel located about 400 m underground at the Olkiluoto site, which is the proposed site for a spent fuel repository in Finland. In the KBS-3H design alternative, each canister, with a surrounding layer of bentonite clay, is pre-packaged in a perforated steel cylinder prior to emplacement in the deposition drift; the entire assembly is called the supercontainer. Several supercontainers are positioned along parallel, 100-300 m long deposition drifts, which are sealed following waste emplacement using drift end plugs. Bentonite distance blocks separate the supercontainers, one from another, along the drift. Steel compartment plugs can be used to seal off drift sections with higher inflow, thus isolating the different compartments within the drift. The present report describes the repository evolution in successive time frames, including key uncertainties. The description of evolution starts with the initial conditions at the time of emplacement of the first canisters. The repository evolves through an early, transient phase to a state where evolution is far slower. Particular attention is given to describing the transient phase, since this is where most of the

Safety assessment for a KBS-3H spent nuclear fuel repository at Olkiluoto. Evolution report

Energy Technology Data Exchange (ETDEWEB)

Smith, Paul; Johnson, Lawrence; Snellman, Margit; Pastina, Barbara; Gribi, Peter

2008-01-15

The KBS-3 method, based on multiple barriers, is the proposed spent fuel disposal method both in Sweden and Finland. KBS-3H and KBS-3V are the two design alternatives of the KBS-3 method. Posiva and SKB have conducted a joint research, demonstration and development (RDandD) programme in 2002-2007 with the overall aim of establishing whether KBS-3H represents a feasible alternative to the reference alternative KBS-3V. The overall objectives of the present phase covering the period 2004-2007, have been to demonstrate that the horizontal deposition alternative is technically feasible and to demonstrate that it fulfils the same long-term safety requirements as KBS-3V. The safety studies conducted as part of this programme include a safety assessment of a preliminary design of a KBS-3H repository for spent nuclear fuel located about 400 m underground at the Olkiluoto site, which is the proposed site for a spent fuel repository in Finland. In the KBS-3H design alternative, each canister, with a surrounding layer of bentonite clay, is pre-packaged in a perforated steel cylinder prior to emplacement in the deposition drift; the entire assembly is called the supercontainer. Several supercontainers are positioned along parallel, 100-300 m long deposition drifts, which are sealed following waste emplacement using drift end plugs. Bentonite distance blocks separate the supercontainers, one from another, along the drift. Steel compartment plugs can be used to seal off drift sections with higher inflow, thus isolating the different compartments within the drift. The present report describes the repository evolution in successive time frames, including key uncertainties. The description of evolution starts with the initial conditions at the time of emplacement of the first canisters. The repository evolves through an early, transient phase to a state where evolution is far slower. Particular attention is given to describing the transient phase, since this is where most of the

Deep repository for spent nuclear fuel. SR-97-Post-closure safety. Main Report. Volume I and II

International Nuclear Information System (INIS)

Hedin, A.

1999-11-01

In preparation for coming site investigations for siting of a deep repository for spent nuclear fuel, the Swedish Government and nuclear regulatory authorities have requested an assessment of the repository's long-term safety. The purpose is to demonstrate whether the risk of harmful effects in individuals in the vicinity of the repository complies with the acceptance criterion formulated by the Swedish regulatory authorities, i.e. that the risk may not exceed 10 -6 per year. Geological data are taken from three sites in Sweden to shed light on different conditions in Swedish granitic bedrock. The future evolution of the repository system is analyzed in the form of five scenarios. The first is a base scenario where the repository is postulated to be built entirely according to specifications and where present-day conditions in the surroundings are postulated to persist. The four other scenarios show how the evolution of the repository differs from that in the base scenario if the repository contains a few initially defective canisters, in the event of climate change, earthquakes, and future inadvertent human intrusion. The time horizon for the analyses is at most one million years, in accordance with preliminary regulations. By means of model studies and calculations, the base scenario analyzes how the radioactivity of the fuel declines with time, the repository's thermal evolution as a result of the decay heat in the fuel, the hydraulic evolution in buffer and backfill when they become saturated with water, and the long-term groundwater flow in the geosphere on the three sites. The overall conclusion of the analyses in the base scenario is that the copper canisters isolating capacity is not threatened by either the mechanical or chemical stresses to which it is subjected. The safety margins are great even in a million-year perspective. The internal evolution in initially defective canisters and the possible resultant migration of radionuclides in buffer, geosphere

The safety case for a HLW repository in Opalinus clay: aims, methodology, first results

International Nuclear Information System (INIS)

Zuidema, Piet

2002-01-01

Piet Zuidema (Nagra, Switzerland) described the development of the safety case for a high level waste repository in Opalinus clay in which canisters would be placed in large vaults. The current phase of work was concerned with demonstrating the feasibility of the disposal concept. The Safety Case is taken to mean a set of arguments to support a statement that the proposed facility will meet relevant safety criteria and will include arguments giving the basis for confidence that those arguments are correct and properly taking account of uncertainties. The safety strategy was concerned both with the inherent robustness of the disposal concept and the adequacy of the assessment capability. As regards the former, the arguments being advanced were primarily qualitative. Key issues in terms of the documentation of the Safety Case were traceability and transparency of information, including how to ensure that key arguments did not become obscured because of the need to make available very large quantities of information

Safety assessment for a KBS-3H spent nuclear fuel repository at Olkiluoto. Summary report

International Nuclear Information System (INIS)

Smith, Paul; Neall, Fiona; Snellman, Margit; Pastina, Barbara; Nordman, Henrik; Johnson, Lawrence; Hjerpe, Thomas

2008-03-01

The KBS-3 method, based on multiple barriers, is the proposed spent fuel disposal method both in Sweden and Finland. KBS-3H and KBS-3V are the two design alternatives of the KBS-3 spent fuel disposal method. Posiva and SKB have conducted a joint research, demonstration and development (RDandD) programme in 2002-2007 with the overall aim of establishing whether KBS-3H represents a feasible alternative to the reference alternative KBS-3V. The overall objectives of the present phase covering the period 2004-2007 have been to demonstrate that the horizontal deposition alternative is technically feasible and to demonstrate that it fulfils the same long-term safety requirements as KBS-3V. The safety studies conducted as part of this programme include a safety assessment of a preliminary design of a KBS-3H repository for spent nuclear fuel located about 400 m underground at the Olkiluoto site, which is the proposed site for a spent fuel repository in Finland. This safety assessment is summarised in the present report. The scientific basis of the safety assessment includes around 30 years of scientific RandD and technical development in the Swedish and Finnish KBS-3V programmes. Much of this scientific basis is directly applicable to KBS-3H. This has allowed the KBS-3H safety studies to focus on those issues that are unique to this design alternative, identified in a systematic 'difference analysis' of KBS-3H and KBS-3V. This difference analysis has shown that the key differences in the evolution and performance of KBS-3H and KBS-3V relate mainly to the engineered barrier system and to the impact of local variations in the rate of groundwater inflow on buffer saturation along the KBS-3H deposition drifts. No features or processes specific to KBS-3H have been identified that could lead to a loss or substantial degradation of the safety functions of the engineered barriers over a million year time frame. Radionuclide release from the repository near field in the event of

Safety assessment for a KBS-3H spent nuclear fuel repository at Olkiluoto. Summary report

Energy Technology Data Exchange (ETDEWEB)

Smith, Paul; Neall, Fiona; Snellman, Margit; Pastina, Barbara; Nordman, Henrik; Johnson, Lawrence; Hjerpe, Thomas

2008-03-15

The KBS-3 method, based on multiple barriers, is the proposed spent fuel disposal method both in Sweden and Finland. KBS-3H and KBS-3V are the two design alternatives of the KBS-3 spent fuel disposal method. Posiva and SKB have conducted a joint research, demonstration and development (RDandD) programme in 2002-2007 with the overall aim of establishing whether KBS-3H represents a feasible alternative to the reference alternative KBS-3V. The overall objectives of the present phase covering the period 2004-2007 have been to demonstrate that the horizontal deposition alternative is technically feasible and to demonstrate that it fulfils the same long-term safety requirements as KBS-3V. The safety studies conducted as part of this programme include a safety assessment of a preliminary design of a KBS-3H repository for spent nuclear fuel located about 400 m underground at the Olkiluoto site, which is the proposed site for a spent fuel repository in Finland. This safety assessment is summarised in the present report. The scientific basis of the safety assessment includes around 30 years of scientific RandD and technical development in the Swedish and Finnish KBS-3V programmes. Much of this scientific basis is directly applicable to KBS-3H. This has allowed the KBS-3H safety studies to focus on those issues that are unique to this design alternative, identified in a systematic 'difference analysis' of KBS-3H and KBS-3V. This difference analysis has shown that the key differences in the evolution and performance of KBS-3H and KBS-3V relate mainly to the engineered barrier system and to the impact of local variations in the rate of groundwater inflow on buffer saturation along the KBS-3H deposition drifts. No features or processes specific to KBS-3H have been identified that could lead to a loss or substantial degradation of the safety functions of the engineered barriers over a million year time frame. Radionuclide release from the repository near field in the

Nuclide transport models for HLW repository safety assessment in Finland, Japan, Sweden, and Canada

Energy Technology Data Exchange (ETDEWEB)

Lee, Young Myoung; Kang, Chul Hyung; Hwang, Yong Soo; Choi, Jong Won; Kim, Sung Gi; Koh, Won Il

1997-10-01

Disposal and design concepts in such countries as Sweden, Finland, Canada and Japan which have already published safety assessment reports for the HLW repositories have been reviewed mainly in view of nuclide transport models used in their assessment. This kind of review would be very helpful in doing similar research in Korea where research program regarding HLW has been just started. (author). 44 refs., 2 tabs., 30 figs

Long term safety requirements and safety indicators for the assessment of underground radioactive waste repositories

International Nuclear Information System (INIS)

Vovk, Ivan

1998-01-01

This presentation defines: waste disposal, safety issues, risk estimation; describes the integrated waste disposal process including quality assurance program. Related to actinides inventory it shows the main results of calculated activity obtained by deterministic estimation. It includes the Radioactive Waste Safety Standards and requirements; features related to site, design and waste package characteristics, as technical long term safety criteria for radioactive waste disposal facilities. Fundamental concern regarding the safety of radioactive waste disposal systems is their radiological impact on human beings and the environment. Safety requirements and criteria for judging the level of safety of such systems have been developed and there is a consensus among the international community on their basis within the well-established system of radiological protection. So far, however, little experience has been gained in applying long term safety criteria to actual disposal systems; consequently, there is an international debate on the most appropriate nature and form of the criteria to be used, taking into account the uncertainties involved. Emerging from the debate is the increasing conviction that the combined use of a variety of indicators would be advantageous in addressing the issue of reasonable assurance in the different time frames involved and in supporting the safety case for any particular repository concept. Indicators including risk, dose, radionuclide concentration, transit time, toxicity indices, fluxes at different points within the system, and barrier performance have all been identified as potentially relevant. Dose and risk are the indicators generally seen as most fundamental, as they seek directly to describe the radiological impact of a disposal system, and these are the ones that have been incorporated into most national standards to date. There are, however, certain problems in applying them. Application of a variety of different indicators

Safety analysis of spent fuel packaging

International Nuclear Information System (INIS)

Akamatsu, Hiroshi; Taniuchi, Hiroaki; Tai, Hideto

1987-01-01

Many types of spent fuel packagings have been manufactured and been used for transport of spent fuels discharged from nuclear power plant. These spent fuel packagings need to be assesed thoroughly about safety transportation because spent fuels loaded into the packaging have high radioactivity and generation of heat. This paper explains the outline of safety analysis of a packaging, Safety analysis is performed for structural, thermal, containment, shielding and criticality factors, and MARC-CDC, TRUMP, ORIGEN, QAD, ANISN, KENO, etc computer codes are used for such analysis. (author)

Use of safety analysis to site comfirmation procedure in case of hard rock repository

International Nuclear Information System (INIS)

Peltonen, E.K.

1984-02-01

The role of safety analysis in a confirmation procedure of a candidate disposal site of radioactive wastes is discussed. Items dealt with include principle reasons and practical goals of the use of safety analysis, methodology of safety analysis and assessment, as well as usefulness and adequacy of the present safety analysis. Safety analysis is a tool, which enables one to estimate quantitatively the possible radiological impacts from the disposal. The results can be compared with the criteria and the suitability conclusions drawn. Because of its systems analytical nature safety analysis is an effective method to reveal, what are the most important factors of the disposal system and the most critical site characteristics inside the lumped parameters often provided by the experimental site investigation methods. Furthermore it gives information on the accuracy needs of different site properties. This can be utilized to judge whether the quality and quantity of the measurements for the characterization are sufficient as well as to guide the further site investigations. A more practical discussion regarding the applicability of the use of safety analysis is presented by an example concerning the assessment of a Finnish candidate site for low- and intermediate-level radioactive waste repository. (author)

Safety Assessment Context for Croatian Low and Intermediate Level Radioactive Waste Repository

International Nuclear Information System (INIS)

Levanat, I.; Lokner, V.

1998-01-01

Safety assessments in a small country are usually performed to support the national waste management strategy, demonstrating compliance with national regulation for a particular facility. However, this assessment should - quite generally - provide reasonable assurance both to the public and to decision makers than the Croatian share of LILW from NPP Krsko can be safely disposed in Croatia. More specifically, assessment should clearly present all realistic options and compare the associated long term repository performances, demonstrating that desirable safety goals can be archived by an appropriate choice of (a) location, (b) facility design, (c) institutional control period and (d) waste acceptance criteria. As relevant national legislation is presently under review, generally recognized international safety standards, criteria and recommendations (e.g. as presented in the recent IAEA publications) should provide guidance for the assessment evaluation, since it is expected that they will be incorporated in the new national regulations. Finally, since Croatian radioactive waste management strategy is yet to be developed, such an assessment may contribute to its formulation and facilitate some specific decisions. (author)

Upgrading of radon's type near surface repository in Latvia

International Nuclear Information System (INIS)

Abramenkovs, A.

2006-01-01

In 1959, the Soviet government decided to construct the near surface radioactive wastes repository 'Radons' near the Baldone city. It was put in operation in 1962. The changes in the development of the repository were induced by the necessarily to upgrade it for disposal of radioactive wastes from the decommissioning of the Salaspils Research Reactor (SRR). The safety assessment of repository was performed during 2000-2001 under the PHARE project for necessary upgrades of repository. The outline design for new vaults and interim storage for long lived radioactive wastes was elaborated during 2003-2004 years. The Environmental Impact Assessment (EIA) for upgrade of Baldone repository was performed during 2004-2005 years. It was found, that additional efforts must be devoted for solution of social aspects o for successful operation and upgrade of repository. It was shown by EIA, that the local population has a negative opinion against the upgrade of repository in Latvia. The main recommendations for upgrades were connected with increasing the safety of repository, increasing of PR activities for education of society and developing of compensation mechanism for local municipality. (author)

Topical session proceedings of the 4. IGSC meeting on: the potential impacts on repository safety from potential partitioning and transmutation programme

International Nuclear Information System (INIS)

Wollrath, Juergen; Voinis, Sylvie; Hadermann, Joerg; Van Luik, Abraham E.

2003-01-01

The bulk of radioactive waste results mainly from energy production in nuclear power plants. At present, this waste is safely stored near reactor sites, in dedicated storage facilities, and low-level wastes in some countries are disposed of in near surface disposal facilities. The stored wastes are accumulating, and the generally agreed-to solution is to dispose of vitrified waste or spent fuel in deep geological disposal facilities. The main concern in the disposal of radioactive waste is related to long-lived radionuclides - some of them will remain hazardous for tens of thousands of years and longer. To demonstrate the safety of geological disposal of radioactive waste a safety case has to be developed for a specific concept of disposal at a given site. This safety case is a collection of arguments at a given stage of repository development in support of the long-term safety of the repository. The safety case comprises the findings of a safety assessment and a statement of confidence in these findings. The type of waste to be disposed of is an important component of the disposal concept. At present, novel fuel cycles are under investigation at national and international levels including research in partitioning and transmutation (P and T) technologies. Comprehensive projects are carried out in OECD countries, notably France, Japan and the USA. Within the Nuclear Fission Safety programme the European Commission (EC) is supporting several projects and it is anticipated that the contribution of the EC will increase within the 6. Framework Programme. In addition the Nuclear Energy Agency has recently published a report comparing fuel cycles for accelerator driven systems and fast reactors. By using P and T it might be possible to reduce the long-lived component of the radioactive waste, thus easing the waste management problem. A safety case for a repository for wastes from P and T fuel cycles would be different from a safety case that deals with spent fuel and

Spent nuclear fuel for disposal in the KBS-3 repository

Energy Technology Data Exchange (ETDEWEB)

Grahn, Per; Moren, Lena; Wiborgh, Maria

2010-12-15

The report is included in a set of Production reports, presenting how the KBS-3 repository is designed, produced and inspected. The set of reports is included in the safety report for the KBS-3 repository and repository facility. The report provides input to the assessment of the long-term safety, SR-Site as well as to the operational safety report, SR-Operation. The report presents the spent fuel to be deposited, and the requirements on the handling and selection of fuel assemblies for encapsulation that follows from that it shall be deposited in the KBS-3 repository. An overview of the handling and a simulation of the encapsulation and the resulting canisters to be deposited are presented. Finally, the initial state of the encapsulated spent nuclear fuel is given. The initial state comprises the radionuclide inventory and other data required for the assessment of the long-term safety

Spent nuclear fuel for disposal in the KBS-3 repository

International Nuclear Information System (INIS)

Grahn, Per; Moren, Lena; Wiborgh, Maria

2010-12-01

The report is included in a set of Production reports, presenting how the KBS-3 repository is designed, produced and inspected. The set of reports is included in the safety report for the KBS-3 repository and repository facility. The report provides input to the assessment of the long-term safety, SR-Site as well as to the operational safety report, SR-Operation. The report presents the spent fuel to be deposited, and the requirements on the handling and selection of fuel assemblies for encapsulation that follows from that it shall be deposited in the KBS-3 repository. An overview of the handling and a simulation of the encapsulation and the resulting canisters to be deposited are presented. Finally, the initial state of the encapsulated spent nuclear fuel is given. The initial state comprises the radionuclide inventory and other data required for the assessment of the long-term safety

Operational safety and radiation protection considerations in designing an HLW repository in Germany

International Nuclear Information System (INIS)

Filbert, W.; Kreienmeyer, M.; Poehler, M.; Niehues, N.

2008-01-01

In Germany the reference concept for disposal of heat generating radioactive waste considers emplacing canisters with vitrified waste in deep vertical boreholes drilled from the drifts of a repository mine in salt at a depth of 870 m. Spent fuel is to be disposed of in self-shielding POLLUX casks in horizontal drifts. An optimized disposal concept anticipates emplacing unshielded canisters with vitrified HLW and canisters containing the fuel rods of 3 PWR or 9 BWR fuel assemblies in boreholes with a diameter of 60 cm and a depth of up to 300 m.. In all cases the void space between POLLUX cask and drifts and canisters and borehole wall will be backfilled with crushed salt. (1) Operational Safety: Based on a detailed description of all underground disposal operation steps, the possible impacts on the disposal operations were analysed and the need for further studies determined. The disposal operation steps comprise e.g. rail bound transport from the shaft to the emplacement drift and emplacement process itself. As possible impacts the following occurrences were considered: ventilation failure, power supply failure, rock mechanics impact including cross-section convergence, irregular floor uplift and rock fall, brine and natural gas intrusion, derailing of transport carts and finally internal fire. (2) Radiation Protection: According to the German Atomic Energy Act (AtG), the design, construction and operation of a nuclear site like a final repository has to be licensed by the responsible authority. The Radiological Protection Ordinance and further guidelines i.e. concerning the emission and immission of released radioactive nuclides or the risk analysis of possible failure, build the basis for the licensing procedures. To ensure adequate protection against undue radiation exposure the repository is divided into different radiological protection areas. Generally, the handling of shielded waste packages above und under ground (including all the pathway of transport and

Reference design description for a geologic repository. Revision 02

International Nuclear Information System (INIS)

1999-01-01

This Reference Design Description explains the current design for a potential geologic repository that may be located at Yucca Mountain in Nevada. It describes the proposed design for a surface facility, subsurface repository, and waste packaging; it also presents the current design of the key engineering systems for the final four phases: operations, monitoring, closure, and postclosure. In addition, this Reference Design Description reviews the expected long-term performance of the potential repository. In accordance with current law, this design does not include an interim storage option. This document has six major sections. Section 1 describes the physical layout of the proposed repository. The second section describes the 4-phase evolution of the development of the proposed repository. Section 3 describes the reception of waste from offsite locations. The fourth section details the various systems that will package the waste and move it below ground as well as safety monitoring and closure. Section 5 describes the systems (both physical and analytical) that ensure continued safety after closure. The final section offers design options that may be adopted to increase the margin of safety

Development of IFC based fire safety assesment tools

DEFF Research Database (Denmark)

Taciuc, Anca; Karlshøj, Jan; Dederichs, Anne

2016-01-01

Due to the impact that the fire safety design has on the building's layout and on other complementary systems, as installations, it is important during the conceptual design stage to evaluate continuously the safety level in the building. In case that the task is carried out too late, additional...... changes need to be implemented, involving supplementary work and costs with negative impact on the client. The aim of this project is to create a set of automatic compliance checking rules for prescriptive design and to develop a web application tool for performance based design that retrieves data from...... Building Information Models (BIM) to evacuate the safety level in the building during the conceptual design stage. The findings show that the developed tools can be useful in AEC industry. Integrating BIM from conceptual design stage for analyzing the fire safety level can ensure precision in further...

Simulating Earthquake Rupture and Off-Fault Fracture Response: Application to the Safety Assessment of the Swedish Nuclear Waste Repository

KAUST Repository

Falth, B.; Hokmark, H.; Lund, B.; Mai, Paul Martin; Roberts, R.; Munier, R.

2014-01-01

To assess the long-term safety of a deep repository of spent nuclear fuel, upper bound estimates of seismically induced secondary fracture shear displacements are needed. For this purpose, we analyze a model including an earthquake fault, which

Safety analysis SFR 1. Long-term safety

Energy Technology Data Exchange (ETDEWEB)

2008-12-15

An updated assessment of the long-term safety of SKB's final repository for radioactive operational waste, SFR 1, is presented in this report. The report is included in the safety analysis report for SFR 1. The most recent account of long-term safety was submitted to the regulatory authorities in 2001. The present report has been compiled on SKB's initiative to address the regulatory authorities' viewpoints regarding the preceding account of long-term safety. Besides the new mode of working with safety functions there is another important difference between the 2001 safety assessment and the current assessment: The time horizon in the current assessment has been extended to 100,000 years in order to include the effect of future climate changes. The purpose of this renewed assessment of the long-term safety of SFR 1 is to show with improved data that the repository is capable of protecting human health and the environment against ionizing radiation in a long-term perspective. This is done by showing that calculated risks lie below the risk criteria stipulated by the regulatory authorities. SFR 1 is built to receive, and after closure serve as a passive repository for, low. and intermediate-level radioactive waste. The disposal chambers are situated in rock beneath the sea floor, covered by about 60 metres of rock. The underground part of the facility is reached via two tunnels whose entrances are near the harbour. The repository has been designed so that it can be abandoned after closure without further measures needing to be taken to maintain its function. The waste in SFR 1 is short-lived low- and intermediate-level waste. After 100 years the activity is less than half, and after 1,000 years only about 2% of the original activity remains. The report on long-term safety comprises eleven chapters. Chapter 1 Introduction. The chapter describes the purpose, background, format and contents of SAR-08, applicable regulations and injunctions, and the regulatory

Safety analysis SFR 1. Long-term safety

International Nuclear Information System (INIS)

2008-12-01

An updated assessment of the long-term safety of SKB's final repository for radioactive operational waste, SFR 1, is presented in this report. The report is included in the safety analysis report for SFR 1. The most recent account of long-term safety was submitted to the regulatory authorities in 2001. The present report has been compiled on SKB's initiative to address the regulatory authorities' viewpoints regarding the preceding account of long-term safety. Besides the new mode of working with safety functions there is another important difference between the 2001 safety assessment and the current assessment: The time horizon in the current assessment has been extended to 100,000 years in order to include the effect of future climate changes. The purpose of this renewed assessment of the long-term safety of SFR 1 is to show with improved data that the repository is capable of protecting human health and the environment against ionizing radiation in a long-term perspective. This is done by showing that calculated risks lie below the risk criteria stipulated by the regulatory authorities. SFR 1 is built to receive, and after closure serve as a passive repository for, low. and intermediate-level radioactive waste. The disposal chambers are situated in rock beneath the sea floor, covered by about 60 metres of rock. The underground part of the facility is reached via two tunnels whose entrances are near the harbour. The repository has been designed so that it can be abandoned after closure without further measures needing to be taken to maintain its function. The waste in SFR 1 is short-lived low- and intermediate-level waste. After 100 years the activity is less than half, and after 1,000 years only about 2% of the original activity remains. The report on long-term safety comprises eleven chapters. Chapter 1 Introduction. The chapter describes the purpose, background, format and contents of SAR-08, applicable regulations and injunctions, and the regulatory

Safety analysis SFR 1. Long-term safety

Energy Technology Data Exchange (ETDEWEB)

2008-12-15

An updated assessment of the long-term safety of SKB's final repository for radioactive operational waste, SFR 1, is presented in this report. The report is included in the safety analysis report for SFR 1. The most recent account of long-term safety was submitted to the regulatory authorities in 2001. The present report has been compiled on SKB's initiative to address the regulatory authorities' viewpoints regarding the preceding account of long-term safety. Besides the new mode of working with safety functions there is another important difference between the 2001 safety assessment and the current assessment: The time horizon in the current assessment has been extended to 100,000 years in order to include the effect of future climate changes. The purpose of this renewed assessment of the long-term safety of SFR 1 is to show with improved data that the repository is capable of protecting human health and the environment against ionizing radiation in a long-term perspective. This is done by showing that calculated risks lie below the risk criteria stipulated by the regulatory authorities. SFR 1 is built to receive, and after closure serve as a passive repository for, low. and intermediate-level radioactive waste. The disposal chambers are situated in rock beneath the sea floor, covered by about 60 metres of rock. The underground part of the facility is reached via two tunnels whose entrances are near the harbour. The repository has been designed so that it can be abandoned after closure without further measures needing to be taken to maintain its function. The waste in SFR 1 is short-lived low- and intermediate-level waste. After 100 years the activity is less than half, and after 1,000 years only about 2% of the original activity remains. The report on long-term safety comprises eleven chapters. Chapter 1 Introduction. The chapter describes the purpose, background, format and contents of SAR-08, applicable regulations and injunctions, and the

Understanding the evolution of the repository and the olkiluoto site

International Nuclear Information System (INIS)

Koskinen, K.; Pastina, B.

2008-01-01

Posiva Safety Case is organised in a portfolio including ten main reports: Site, Spent Fuel Characteristics and Inventories, Canister Design, Repository Design, Process, Evolution of the Repository and the Site, Biosphere Assessment, Radionuclide Transport, Complementary Evaluations of Safety, and Summary. This portfolio constitutes the basis of the Preliminary Safety Assessment Report, which will be presented to the authorities in 2012 as part of the repository construction license application. The Evolution report [1], which is the focus of this paper, is the main advance in the Safety Case portfolio since the implementation of the Safety Case plan [2] in 2005. The report provides the status of current knowledge with respect to the evolution of the site and the engineered barrier system and highlights areas where better understanding is needed. (authors)

Proceedings of the workshop on radionuclide release scenarios for geologic repositories

Energy Technology Data Exchange (ETDEWEB)

1981-01-01

The safety of radioactive waste disposal in geological formations cannot be verified experimentally. Safety analysis provides the only means to ensure that all risks associated with the waste repositories are acceptably low. The definition of radionuclide release scenarios, as discussed in these proceeedings, is the first step in the safety analysis of waste repositories.

Repository operational criteria analysis

International Nuclear Information System (INIS)

Hageman, J.P.; Chowdhury, A.H.

1992-08-01

The objective of the ''Repository Operational Criteria (ROC) Feasibility Studies'' (or ROC task) was to conduct comprehensive and integrated analyses of repository design, construction, and operations criteria in 10 CFR Part 60 regulations, considering the interfaces and impacts of any potential changes to those regulations. The study addresses regulatory criteria related to the preclosure aspects of the geologic repository. The study task developed regulatory concepts or potential repository operational criteria (PROC) based on analysis of a repository's safety functions and other regulations for similar facilities. These regulatory concepts or PROC were used as a basis to assess the sufficiency and adequacy of the current criteria in 10 CFR Part 60. Where the regulatory concepts were same as current operational criteria, these criteria were referenced. The operations criteria referenced or the PROC developed are given in this report. Detailed analyses used to develop the regulatory concepts and any necessary PROC for those regulations that may require a minor change are also presented. The results of the ROC task showed a need for further analysis and possible major rule change related to the design bases of a geologic repository operations area, siting, and radiological emergency planning

Deep repository for spent nuclear fuel. SR 97 - Post-closure safety. Main Report. Summary

Energy Technology Data Exchange (ETDEWEB)

Hedin, A. [ed.

1999-11-01

In preparation for coming site investigations for siting of a deep repository for spent nuclear fuel, the Swedish Government and nuclear regulatory authorities have requested an assessment of the repository's long-term safety. The purpose is to demonstrate whether the risk of harmful effects in individuals in the vicinity of the repository complies with the acceptance criterion formulated by the Swedish regulatory authorities, i.e. that the risk may not exceed 10{sup -6} per year. Geological data are taken from three sites in Sweden to shed light on different conditions in Swedish granitic bedrock. The future evolution of the repository system is analyzed in the form of five scenarios. The first is a base scenario where the repository is postulated to be built entirely according to specifications and where present-day conditions in the surroundings are postulated to persist. The four other scenarios show how the evolution of the repository differs from that in the base scenario if the repository contains a few initially defective canisters, in the event of climate change, earthquakes, and future inadvertent human intrusion. The time horizon for the analyses is at most one million years, in accordance with preliminary regulations. By means of model studies and calculations, the base scenario analyzes how the radioactivity of the fuel declines with time, the repository's thermal evolution as a result of the decay heat in the fuel, the hydraulic evolution in buffer and backfill when they become saturated with water, and the long-term groundwater flow in the geosphere on the three sites. The overall conclusion of the analyses in the base scenario is that the copper canisters isolating capacity is not threatened by either the mechanical or chemical stresses to which it is subjected. The safety margins are great even in a million-year perspective. The internal evolution in initially defective canisters and the possible resultant migration of radionuclides in

Continuous Improvement and the Safety Case for the Waste Isolation Pilot Plant Geologic Repository - 13467

Energy Technology Data Exchange (ETDEWEB)

Van Luik, Abraham; Patterson, Russell; Nelson, Roger [US Department of Energy, Carlsbad Field Office, 4021 S. National parks Highway, Carlsbad, NM 88220 (United States); Leigh, Christi [Sandia National Laboratories Carlsbad Operations, 4100 S. National parks Highway, Carlsbad, NM 88220 (United States)

2013-07-01

The Waste Isolation Pilot Plant (WIPP) is a geologic repository 2150 feet (650 m) below the surface of the Chihuahuan desert near Carlsbad, New Mexico. WIPP permanently disposes of transuranic waste from national defense programs. Every five years, the U.S. Department of Energy (DOE) submits an application to the U.S. Environmental Protection Agency (EPA) to request regulatory-compliance re-certification of the facility for another five years. Every ten years, DOE submits an application to the New Mexico Environment Department (NMED) for the renewal of its hazardous waste disposal permit. The content of the applications made by DOE to the EPA for re-certification, and to the NMED for permit-renewal, reflect any optimization changes made to the facility, with regulatory concurrence if warranted by the nature of the change. DOE points to such changes as evidence for its having taken seriously its 'continuous improvement' operations and management philosophy. Another opportunity for continuous improvement is to look at any delta that may exist between the re-certification and re-permitting cases for system safety and the consensus advice on the nature and content of a safety case as being developed and published by the Nuclear Energy Agency's Integration Group for the Safety Case (IGSC) expert group. DOE at WIPP, with the aid of its Science Advisor and teammate, Sandia National Laboratories, is in the process of discerning what can be done, in a reasonably paced and cost-conscious manner, to continually improve the case for repository safety that is being made to the two primary regulators on a recurring basis. This paper will discuss some aspects of that delta and potential paths forward to addressing them. (authors)

Investigations of possibilities to dispose of spent nuclear fuel in Lithuania: a model case. Volume 3, Generic Safety Assessment of Repository in Crystalline Rocks

International Nuclear Information System (INIS)

Motiejunas, S.; Poskas, P.

2005-01-01

In this Volume a generic safety assessment of the repository for spent nuclear fuel in crystalline rock in Lithuania is presented. Modeling of safety relevant radionuclide release from the defected canister and their transport through the near field and far field was performed. Doses to humans due to released radionuclides in the well water were calculated and compared with the dose restrictions existing in Lithuania. For this stage of generic safety assessment only two scenarios were chosen: base scenario and canister defect scenario. KBS-3 concept developed by SKB for disposal of spent nuclear fuel in Sweden was chosen as prototype for repository in crystalline basement in Lithuania. The KBS-3H design with horizontal canister emplacement is proposed as a reference design for Lithuania

Design aspects of the alpha repository. II. Conceptual layouts of underground storage facilities

International Nuclear Information System (INIS)

Grams, W.H.

1975-01-01

Five conceptual repository layouts are presented: linear repository, 1 panel; bow tie repository, 2 panels; maltese cross repository, 4 panels; snowflake repository; 5 panels, and sash window repository, 8 panels. The layouts are compared with respect to excavation requirements, haulage distances, ventilation flow path designs, and safety features

Evaluation of geological documents available for provisional safety analyses of potential sites for nuclear waste repositories - Are additional geological investigations needed?

International Nuclear Information System (INIS)

2010-10-01

The procedure for selecting repository sites for all categories of radioactive waste in Switzerland is defined in the conceptual part of the Sectoral Plan for Deep Geological Repositories, which foresees a selection of sites in three stages. In Stage I, Nagra proposed geological siting regions based on criteria relating to safety and engineering feasibility. The Swiss Government (the Federal Council) is expected to decide on the siting proposals in 2011. The objective of Stage 2 is to prepare proposals for the location of the surface facilities within the planning perimeters defined by the Federal Council in its decision on Stage 1 and to identify potential sites. Nagra also has to carry out a provisional safety analysis for each site and a safety-based comparison of the sites. Based on this, and taking into account the results of the socio-economic-ecological impact studies, Nagra then has to propose at least two sites for each repository type to be carried through to Stage 3. The proposed sites will then be investigated in more detail in Stage 3 to ensure that the selection of the sites for the General Licence Applications is well founded. In order to realise the objectives of the upcoming Stage 2, the state of knowledge of the geological conditions at the sites has to be sufficient to perform the provisional safety analyses. Therefore, in preparation for Stage 2, the conceptual part of the Sectoral Plan requires Nagra to clarify the need for additional investigations aimed at providing input for the provisional safety analyses. The purpose of the present report is to document Nagra's technical-scientific assessment of this need. The focus is on evaluating the geological information based on processes and parameters that are relevant for safety and engineering feasibility. In evaluating the state of knowledge the key question is whether additional information could lead to a different decision regarding the selection of the sites to be carried through to Stage 3

Postclosure safety assessment of a deep geological repository for Canada's used nuclear fuel

Energy Technology Data Exchange (ETDEWEB)

Hunt, N.G.; Kremer, E.P.; Garisto, F.; Gierszewski, P.; Gobien, M.; Medri, C.L.D. [Nuclear Waste Management Organization, Toronto, ON (Canada); Avis, J.D. [Geofirma Engineering Ltd., Ottawa, ON (Canada); Chshyolkova, T.; Kitson, C.I.; Melnyk, W.; Wojciechowski, L.C. [Atomic Energy of Canada Limited, Pinawa, MB (Canada)

2011-07-01

This paper reports on elements of a postclosure safety assessment performed for a conceptual design and hypothetical site for a deep geological repository for Canada's used nuclear fuel. Key features are the assumption of a copper used fuel container with a steel inner vessel, container placement in vertical in-floor boreholes, a repository depth of 500 m, and a sparsely fractured crystalline rock geosphere. The study considers a Normal Evolution Scenario together with a series of Disruptive Event Scenarios. The Normal Evolution Scenario is a reasonable extrapolation of present day site features and receptor lifestyles, while the Disruptive Event Scenarios examine abnormal and unlikely failures of the containment and isolation systems. Both deterministic and probabilistic simulations were performed. The results show the peak dose consequences occur far in the future and are well below the applicable regulatory acceptance criteria and the natural background levels. (author)

Joint SKI and SSI review of SKB preliminary safety assessment of repository for long-lived low- and intermediate-level waste. Review report

Energy Technology Data Exchange (ETDEWEB)

NONE

2001-03-01

SKI and SSI find that SKB's first proper safety assessment of the SFL 3-5 repositories provides a valuable springboard for continued efforts in this field. Even though the safety assessment is relatively limited in scope, it has numerous merits. The specific problems associated with the chosen repository concept for SFL 3-5 are discussed in a generally transparent manner. On the other hand, the authorities consider that SKB have only partly achieved the expressed goal of studying the significance of the current repository design and the choice of site. The greatest deficiency consists in that neither internal disturbances (such as considerable cracking or degradation of concrete structures) nor external disturbances (such as the effects of climate changes and glaciation) have been addressed in a thorough manner. A coherent report justifying the design choice from a long-term safety perspective is, in large part, not found here. SKI and SSI recommend that SKB provide a comparison with other possible SFL 3-5 repository designs. Depending upon, among other factors, what geospheric and biospheric conditions are assumed, SKB have shown that the calculated dose values could be relatively high for certain cases. More realistic assessments would be needed to draw reasonable comparisons between different sites, and to evaluate the importance of different nuclides in different contexts. Our review of SKBs preliminary safety assessment indicates that a great deal of research and development work remains to be done before the level of knowledge in this field is comparable with that associated with the final repository for spent fuel. This is reflected with unanimity in the international expert committee's review, and in the consultants' reviews. SKI and SSI wish to point out in particular the fact that comparison with SFR is of limited value, since the safety associated with SFL 3- 5 must be assessed on a much longer time scale. SKI and SSI find it remarkable

Multibarrier system preventing migration of radionuclides from radioactive waste repository

Directory of Open Access Journals (Sweden)

Olszewska Wioleta

2015-09-01

Full Text Available Safety of radioactive waste repositories operation is associated with a multibarrier system designed and constructed to isolate and contain the waste from the biosphere. Each of radioactive waste repositories is equipped with system of barriers, which reduces the possibility of release of radionuclides from the storage site. Safety systems may differ from each other depending on the type of repository. They consist of the natural geological barrier provided by host rocks of the repository and its surroundings, and an engineered barrier system (EBS. The EBS may itself comprise a variety of sub-systems or components, such as waste forms, canisters, buffers, backfills, seals and plugs. The EBS plays a major role in providing the required disposal system performance. It is assumed that the metal canisters and system of barriers adequately isolate waste from the biosphere. The evaluation of the multibarrier system is carried out after detailed tests to determine its parameters, and after analysis including mathematical modeling of migration of contaminants. To provide an assurance of safety of radioactive waste repository multibarrier system, detailed long term safety assessments are developed. Usually they comprise modeling of EBS stability, corrosion rate and radionuclide migration in near field in geosphere and biosphere. The principal goal of radionuclide migration modeling is assessment of the radionuclides release paths and rate from the repository, radionuclides concentration in geosphere in time and human exposure to ionizing radiation

Safety assesment on radioactive waste from the partitioning and transmutation fuel cycle

International Nuclear Information System (INIS)

Lee, Youn Myoung; Hwang, Yong Soo; Kang, Chul Hyung; Kim, Sung Gi; Park, Won Suk

2000-12-01

A preliminary study on the quantitative effect of the partition and transmutation on the permanent disposal of HLW, which means the spent fuel in view of current Korean situation, was carried out. Two approaches in quantitative way are considered to be available for evaluating the deterministic influence of P and T strategy on the long-term disposal of this HLW are assessments of waste toxicity indices (TIs) and the repository performance assessments (PAs). TI is measures of the intrinsic radiotoxicity of the wastes and does not incorporate any detailed consideration of the feature, event and processes (FEPs) which might be lead to the release of the nuclides from the waste disposed of in the repository and the transport to and through the biosphere. Whereas, PA, which treated as main topic of present study, does include consideration of such FEPs even though it could not fully comprehensive at the current stage of R andD on geological disposal. Through the study, after reviewing the PA approaches which considered by some countries, relative advantages in case P and T will be performed before disposal over direct permanent disposal. Even though P and T could be an ideal solution to reduce the inventory which eventually decreases the release time as well as the peaks in the annual dose and even minimize the repository area through the proper handling of nuclides whose decay heat is significant and further produce the electricity, it should overcome the such major disadvantages as problems technically exposed during developing and improving the P and T system, economic point of view, and public acceptance in view of environment-friendly issues. In this regard some relevant issues are also discussed to show the direction for further studies

The network to review natural analogue studies and their applications to repository safety assessment and public communication (NAnet)

Energy Technology Data Exchange (ETDEWEB)

Miller, W.M.; Hooker, P.J. [ENVIROS Consulting ltd, 61, the Shore Leith, UK-0 EH6 6RA Edinburgh (United Kingdom)

2004-07-01

Analogue information can increase our conceptual understanding of long-term repository behaviour in support of post-closure performance assessment (PA), provide quantitative data for PA models and provide ways of communicating safety information to non-specialist audiences. These functions of analogue studies have, however, received too little attention in PA reports and safety cases. Many analogue studies have been undertaken in the last two decades costing tens of millions of euros, and these have covered a wide range of phenomena such as uranium ore deposition, natural fission reactors, natural nuclide migration, contaminant containment by clays and sediments, preservation of ancient fossil trees and buried artefacts etc. The different uses of analogues would be easier to manage if a single database of quality approved analogue information were to be created. NAnet, a Thematic Network within the 5. EURATOM FP is aiming to promote more considered applications of analogues in performance and safety assessments and in audience dialogue. NAnet intends critically to review a number of analogue studies in terms of their relevance and limitations to different repository concepts and environments and with regard to their applications in performance assessments, safety cases and communication. On the basis of these reviews, a simple digital database is being developed for the PA community which will allow PA modelers to make quicker and wider use of natural analogue information in performance and safety assessments. It is expected that some of these tools will help radioactive waste institutions to make better use of natural analogue information for communication with different audiences, including the public. (authors)

Coupling fuel cycles with repositories: how repository institutional choices may impact fuel cycle design

International Nuclear Information System (INIS)

Forsberg, C.; Miller, W.F.

2013-01-01

The historical repository siting strategy in the United States has been a top-down approach driven by federal government decision making but it has been a failure. This policy has led to dispatching fuel cycle facilities in different states. The U.S. government is now considering an alternative repository siting strategy based on voluntary agreements with state governments. If that occurs, state governments become key decision makers. They have different priorities. Those priorities may change the characteristics of the repository and the fuel cycle. State government priorities, when considering hosting a repository, are safety, financial incentives and jobs. It follows that states will demand that a repository be the center of the back end of the fuel cycle as a condition of hosting it. For example, states will push for collocation of transportation services, safeguards training, and navy/private SNF (Spent Nuclear Fuel) inspection at the repository site. Such activities would more than double local employment relative to what was planned for the Yucca Mountain-type repository. States may demand (1) the right to take future title of the SNF so if recycle became economic the reprocessing plant would be built at the repository site and (2) the right of a certain fraction of the repository capacity for foreign SNF. That would open the future option of leasing of fuel to foreign utilities with disposal of the SNF in the repository but with the state-government condition that the front-end fuel-cycle enrichment and fuel fabrication facilities be located in that state

Presentation of safety after closure of the repository for spent nuclear fuel. Main report of the project SR-Site. Part III

International Nuclear Information System (INIS)

2011-01-01

The purpose of the safety assessment SR-Site is to investigate whether a safe repository for spent nuclear fuel by KBS-3 type can be constructed at Forsmark in Oesthammar in Sweden. The location of the Forsmark has been selected based on results of several surveys from surface conditions at depth in Forsmark and in Laxemar in Oskarshamn. The choice of location is not justified in SR-Site Report, but in other attachments to SKB's permit applications. SR-Site Report is an important part of SKB's permit applications to construct and operate a repository for spent nuclear fuel at Forsmark in Oesthammar. The purpose of the report in the applications is to show that a repository at Forsmark is safe after closure

Presentation of safety after closure of the repository for spent nuclear fuel. Main report of the project SR-Site. Part I

International Nuclear Information System (INIS)

2011-01-01

The purpose of the safety assessment SR-Site is to investigate whether a safe repository for spent nuclear fuel by KBS-3 type can be constructed at Forsmark in Oesthammar in Sweden. The location of the Forsmark has been selected based on results of several surveys from surface conditions at depth in Forsmark and in Laxemar in Oskarshamn. The choice of location is not justified in SR-Site Report, but in other attachments to SKB's permit applications. SR-Site Report is an important part of SKB's permit applications to construct and operate a repository for spent nuclear fuel at Forsmark in Oesthammar. The purpose of the report in the applications is to show that a repository at Forsmark is safe after closure

VerSi - A Methodology for a Comparison of Potential Repository Sites

International Nuclear Information System (INIS)

Hund, Wilhelm

2010-09-01

In the year 2000 the moratorium on the exploration of the Gorleben salt dome as a potential repository for all kinds of radioactive waste became effective as a result of the consensus agreement between the Federal Government and the utilities about phasing out nuclear energy in Germany. All exploration activities were interrupted for at maximum ten years to clarify conceptual and safety relevant questions. A new set of safety requirements for the final disposal of heat-generating radioactive waste in deep geological formations was established in July 2009 by the Federal Ministry for the Environment, Nature Conservation and Nuclear Safety (BMU). As the BMU intended to carry out a comparison of potential repository sites it was necessary to initiate the development of a methodology for the identification of the site with the highest level of safety. A comparison of different repository sites requires a tool ensuring most confident and objective criteria for the comparison, whereas up to present long-term safety analyses were focused on confirming the suitability of sites by meeting the protection objectives by the measures of dose and risk. Within the 2006 established project VerSi a methodology for comparing different sites in different host rocks will be developed on the basis of long-term safety analyses taking into account geoscientific databases, inventory of radioactive waste, waste containers, corresponding disposal concepts and the feasibility of appropriate backfilling and closure concepts. The development of the method is aiming at providing measures other than dose and risk for the evaluation of the level of safety. For testing the tools a HLW-repository hosted in a salt dome (Gorleben) will be compared with a generic HLW-repository in consolidated clay as a host rock. As until now in Germany no clay stone site has been investigated for hosting a HLW repository, the required data are transferred from international research projects and repository concepts

Initialization of Safety Assessment Process for the Croatian Radioactive Waste repository on Trgovska gora

International Nuclear Information System (INIS)

Lokner, V.; Levanat, I.; Subasic, D.

2000-01-01

An iterative process of safety assessment, presently focusing on the site-specific evaluation of the post-closure phase for the prospective LILW repository on Trgovska gora in Croatia, has recently been initiated. The primary aim of the first assessment iterations is to provide the experts involved, the regulators and the general public with a reasonable assurance that the applicable long term performance and safety objectives can be met. Another goal is to develop a sufficient understanding of the system behavior to support decisions about the site investigation, the facility design, the waste acceptance criteria and the closure conditions. In this initial phase, the safety assessment is structured in a manner following closely methodology of the ISAM. The International Programme for Improving Long Term Safety Assessment Methodologies for Near Surface Radioactive Waste Disposal Facilities the IAEA coordinated research program started in 1997. Results of the safety assessment first iteration will be organized and presented in the form of a preliminary safety analysis report (PSAR), expected to be completed in the second part of the year 2000. As the first report on the initiated safety assessment activities, the PSAR will describe the concept and aims of the assessment process. Particular emphasis will be placed on description of the key elements of a safety assessment approach by: a) defining the assessment context; b) providing description of the disposal system; c) developing and justifying assessment scenarios; d) formulating and implementing models; and e) interpreting the scoping calculations. (author)

Evaluation of behaviour and Safety in a geologic deep repository; Evaluacion del comportamiento y de la seguridad de un almacenamiento geologico profundo en granito

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-07-01

This report presents a comprehensive description of the post-closure radiological safety assessment of a repository for the spent fuel arisings resulting from the Spanish nuclear program. This Safety Assessment constitutes a first step within a systematical process that will permit, thorough successive approximations, to predict the performance of the different barriers of the disposal system, and its capability to comply with the assigned safety functions and with the established safety criteria. The primary bases for this Safety Assessment are the following: The disposal concept considers the storage of the fuel assemblies in carbon steel canisters of 10 cm of thickness, emplaced horizontally in galleries excavated in granite of 2,4 m of diameter and 500 m of length, using a bentonite thickness of 75 cm around canisters as buffer material. The repository is located in a granitic site defined with available data about surface characteristics of Spanish granites. The exercise uses a probabilistic approximation in order to cope with the uncertainties associated with the different imputs parameters. (Author)

Safety case for the disposal of spent nuclear fuel at Olkiluoto. Models and data for the repository system 2012. Parts 1 and 2

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-09-15

TURVA-2012 is Posiva Oy's safety case in support of the Preliminary Safety Analysis Report (PSAR 2012) and application for a construction licence for a KBS-3V spent nuclear fuel repository. The present report is a key element of the TURVA-2012 report portfolio and has the objective of documenting the models, data, assumptions and treatment of uncertainties in the context of the safety case. This report is the main link between the safety case and the engineered barrier design and their development as well as between the safety case and the Olkiluoto site investigations. This report focuses on the models and data used in Performance Assessment and in Assessment of Radionuclide Release Scenarios for the Repository System, which are key reports of TURVA-2012. Models and data for the surface environment are discussed in dedicated biosphere modelling and data reports. This report describes the methodology for the identification of key models and data as well as the modelling chain with input and output data connections. Models and data are presented for all components of the repository system: spent nuclear fuel, canister, buffer, backfill, closure, underground openings and geosphere. The report is structured so that the modelling of external processes is discussed first, followed by the models and data used in the performance assessment to address the evolution of the repository system and finally the models and data used in the radionuclide release and transport assessment. Confidence in the models and data and the treatment of uncertainties are also discussed. The present report traces the path from data production to implementation in the modelling chain. During the compilation of the report, some discrepancies between the sources of data and data usage, as well as some inconsistencies in model assumptions, were identified. The consequences of the potentially most significant of these were checked through additional radionuclide release and transport

Safety case for the disposal of spent nuclear fuel at Olkiluoto. Models and data for the repository system 2012. Parts 1 and 2

International Nuclear Information System (INIS)

2013-09-01

TURVA-2012 is Posiva Oy's safety case in support of the Preliminary Safety Analysis Report (PSAR 2012) and application for a construction licence for a KBS-3V spent nuclear fuel repository. The present report is a key element of the TURVA-2012 report portfolio and has the objective of documenting the models, data, assumptions and treatment of uncertainties in the context of the safety case. This report is the main link between the safety case and the engineered barrier design and their development as well as between the safety case and the Olkiluoto site investigations. This report focuses on the models and data used in Performance Assessment and in Assessment of Radionuclide Release Scenarios for the Repository System, which are key reports of TURVA-2012. Models and data for the surface environment are discussed in dedicated biosphere modelling and data reports. This report describes the methodology for the identification of key models and data as well as the modelling chain with input and output data connections. Models and data are presented for all components of the repository system: spent nuclear fuel, canister, buffer, backfill, closure, underground openings and geosphere. The report is structured so that the modelling of external processes is discussed first, followed by the models and data used in the performance assessment to address the evolution of the repository system and finally the models and data used in the radionuclide release and transport assessment. Confidence in the models and data and the treatment of uncertainties are also discussed. The present report traces the path from data production to implementation in the modelling chain. During the compilation of the report, some discrepancies between the sources of data and data usage, as well as some inconsistencies in model assumptions, were identified. The consequences of the potentially most significant of these were checked through additional radionuclide release and transport calculations

Assesment of opportunities for landfill gas utilisation in Bulgaria

International Nuclear Information System (INIS)

Gramatikov, S.; Iliev, I.; Andreev, S.; Hristoskov, I.

2011-01-01

In Bulgaria, about 14 million tons annually of municipal solid waste (MSW) are collected and disposed of in landfills - about 618 kg/capita annually. The implementation of Landfill Gas (LFG) energy recovery/utilization projects in Bulgaria serves as an essential landfill management strategy, and can also reduce greenhouse gases and air pollutants, leading to improved local air quality and reduced health risks. Results of assesment landfill tests of several municipalities, made by the team of Encon Services for estimation of the potential of their sites are shown in this paper. (authors)

International Conference on Geological Repositories 2016. Conference Synthesis, 7-9 December 2016, Paris, France

International Nuclear Information System (INIS)

Walke, Russell; Kwong, Gloria; )

2017-01-01

Worldwide consensus exists within the international community that geological repositories can provide the necessary long-term safety and security to isolate long-lived radioactive waste from the human environment over long timescales. Such repositories are also feasible to construct using current technologies. However, proving the technical merits and safety of repositories, while satisfying societal and political requirements, has been a challenge in many countries. Building upon the success of previous conferences held in Denver (1999), Stockholm (2003), Berne (2007) and Toronto (2012), the ICGR 2016 brought together high-level decision makers from regulatory and local government bodies, waste management organisations and public stakeholder communities to review current perspectives of geological repository development. This publication provides a synthesis of the 2016 conference on continued engagement and safe implementation of repositories, which was designed to promote information and experience sharing, particularly in the development of polices and regulatory frameworks. Repository safety, and the planning and implementation of repository programs with societal involvement, as well as ongoing work within different international organisations, were also addressed at the conference. (authors)

Landscape modeling for dose calculations in the safety assessment of a repository for spent nuclear fuel

International Nuclear Information System (INIS)

Lindborg, Tobias; Kautsky, Ulrik; Brydsten, Lars

2007-01-01

The Swedish Nuclear Fuel and Waste Management Co.,(SKB), pursues site investigations for the final repository for spent nuclear fuel at two sites in the south eastern part of Sweden, the Forsmark- and the Laxemar site. Data from the two site investigations are used to build site descriptive models of the areas. These models describe the bedrock and surface system properties important for designing the repository, the environmental impact assessment, and the long-term safety, i.e. up to 100,000 years, in a safety assessment. In this paper we discuss the methodology, and the interim results for, the landscape model, used in the safety assessment to populate the Forsmark site in the numerical dose models. The landscape model is built upon ecosystem types, e.g. a lake or a mire, (Biosphere Objects) that are connected in the landscape via surface hydrology. Each of the objects have a unique set of properties derived from the site description. The objects are identified by flow transport modeling, giving discharge points at the surface for all possible flow paths from the hypothetical repository in the bedrock. The landscape development is followed through time by using long-term processes e.g. shoreline displacement and sedimentation. The final landscape model consists of a number of maps for each chosen time period and a table of properties that describe the individual objects which constitutes the landscape. The results show a landscape that change over time during 20,000 years. The time period used in the model equals the present interglacial and can be used as an analogue for a future interglacial. Historically, the model area was covered by sea, and then gradually changes into a coastal area and, in the future, into a terrestrial inland landscape. Different ecosystem types are present during the landscape development, e.g. sea, lakes, agricultural areas, forest and wetlands (mire). The biosphere objects may switch from one ecosystem type to another during the

Elements of the safety case for the Morsleben repository based on probabilistic modelling

International Nuclear Information System (INIS)

Wollrath, J.; Niemeyer, M.; Resele, G.; Becker, D.A.; Hirsekorn, P.

2008-01-01

The Morsleben nuclear waste repository (ERAM) for low- and intermediate-level mainly short-lived waste is located in a former salt mine. The closure concept was developed in parallel and interacting with the safety assessment. The safety concept is based on extensive backfilling with salt concrete complemented with seals between the main disposal areas and the rest of the mine building. Thus, the entire system exhibits a barrier effect through a partially redundant combination of several processes. However, in the formal safety assessment no credit is taken from the barrier effect of the extensive backfill. In the safety assessments, the different possibilities of system development, the resulting array of potential fluid movement and a large number of potential radionuclide migration pathways are mapped in the bandwidth of derived parameters. The calculated response of the system to parameter variations is non-linear. Different processes may compete and compensate each other. Hence, the common practice to choose a conservative parameter set for the safety assessment is a priori impossible. The safety assessment has been performed independently by two groups with different computer models, for the same closure concept and the same basic parameters but independent conceptual approaches. Both groups have performed deterministic and probabilistic dose calculations. The results match well; the differences can be explained on basis of the model approaches. Although a large bandwidth is considered for a number of parameters the maximum radiation exposure remains clearly below the applicable dose limit for nearly all calculations, demonstrating the robustness of the system. These aspects significantly contribute to confidence building in the Safety Case for ERAM. (authors)

Evaluation of radiological safety assessment of a repository in a clay rock formation. Evaluacion del comportamiento y de la seguridad de un almacenamiento profundo en arcilla

Energy Technology Data Exchange (ETDEWEB)

1999-12-15

This report presents a comprehensive description of the post-closure radiological safety assessment of a repository for the spent fuel arisings resulting from the Spanish nuclear program excavated in a clay host rock formation. In this report three scenarios have been analysed in detail. The first scenario represents the normal in detail. The first scenario represents the normal evolution of the repository (Reference Scenario); and includes a set of variants to investigate the relative importance of the various repository components and examine the sensitivity of the performance to parameters variations. Two altered scenarios have also been considered: deep well construction and poor sealing of the repository. This document contains a detailed description of the repository system, the methodology adopted for the scenarios generation, the process modelling approach and the results of the consequences analysis. (Author)

'Practicality' as a key constraint to HLW repository design

International Nuclear Information System (INIS)

Kitayama, Kazumi; Sakabe, Yasushi; Ishiguro, Katsuhiko

2007-01-01

Designs of repositories in Japan for HLW have focused very much on demonstration of post-closure safety. Safety can be assured using very simple assessment techniques, which make many conservative simplifications. Such a situation is reasonable for the early stages of generic concept demonstration, but becomes less appropriate as NUMO moves towards siting, where a number of issues involved with construction and operation of a repository - generally grouped together as 'practicality'. The engineering logistics and conventional safety of repository construction and operation have been relatively little studied and present major challenges. Current designs emphasise a minimum of infrastructure in the emplacement tunnels and remote-handled operation. This would be difficult enough, but such operations need to be carried out to strict quality limits and need to be robust in the event of equipment failure or disruptive events. The paper will first examine how designs can be modified from the viewpoint of logistics. The implications of such modifications on operational robustness and associated safety in case of perturbation scenarios are then considered. (author)

Impact of retrievability of repository design

International Nuclear Information System (INIS)

Heijdra, J.J.; Gaag, J. v.d.; Prij, J.

1995-01-01

In this paper the impact of the retrievability on the design of the repository will be handled. Retrievability of radioactive waste from a repository in geological formations has received increasing attention during recent years. It is obvious that this retrievability will have consequences in terms of mining engineering, safety and cost. The purpose of the present study is to evaluate cost consequences by comparing two extreme options for retrievable storage. (author). 6 refs., 3 figs

SR-CAN - a safety assessment of a repository of spent nuclear fuel: canister performance and effects on the biosphere

International Nuclear Information System (INIS)

Kautsky, U.; Kumblad, L.

2004-01-01

During the next few years the Swedish Nuclear Fuel and Waste Management Co. (SKB) performs site investigations at two sites in Sweden for a future repository of spent nuclear fuel. Parallel an encapsulation plant is planned to encapsulate the spent fuel in copper canisters according to the KBS-3 method. The purpose of the SR-CAN safety assessment is to show the performance of the canister isolations at different sites for a repository at 500 meters depth in crystalline rock. Moreover, SR-CAN provides an example how the site specific safety assessment of a deep repository will be made in year 2006-2008. To be able to calculate dose and risk for humans and the environment, new assessment methods were developed for the biosphere. These methods were based on a system ecological approach and used knowledge from landscape ecology to provide an integrated approach with hydrology and geology considering the discharges in a watershed and calculating consequences in terrestrial and aquatic (freshwater and marine) ecosystems. A range of methods and tools were developed in GIS and Matlab/Simulink to be able to model and understand the important processes in the landscape today and during the next few thousands of years. In this paper, an overview of the program and the novel methods are presented, as well as some examples from performance calculations from a watershed in the Forsmark area considering effects on humans and ecosystems. (author)

International perspective on repositories for low level waste

International Nuclear Information System (INIS)

Bergstroem, Ulla; Pers, Karin; Almen, Ylva

2011-12-01

Nuclear energy production gives rise to different types of radioactive waste. The use of nuclear isotopes within the research, industry and medical sectors also generates radioactive waste. To protect man and the environment from radiation the waste is isolated and contained by deposition in repositories. These repositories may have various designs regarding location, barriers etc depending on the potential danger of the waste. In Sweden, low- and intermediate level waste (LILW) is disposed of in the SFR repository in Forsmark. The repository is located 60 metres down into the bedrock under the bottom of the sea and covered by 6 metres of water. It is planned to extend SFR to accommodate decommissioning waste from the dismantling of the Swedish nuclear power facilities and also for the additional operation waste caused by the planned prolonged operation time. When planning the extension consultations will be carried out with the host municipality, authorities, organisations and general public. In planning the extension, SKB has performed a worldwide compilation of how other countries have, or plan to, handle the final disposal of similar wastes. The aim of this report is to give a brief description of LILW repositories worldwide; including general brief descriptions of many facilities, descriptions of the waste and the barriers as well as safety assessments for a few chosen repositories which represent different designs. The latter is performed, where possible, to compare certain features against the Swedish SFR. To provide a background and context to this study, international organisations and conventions are also presented along with internationally accepted principles regarding the management of radioactive waste. Similar to SFR, suitable locations for the repositories have, in many countries, been found at sites that already have, or used to have nuclear activities, such as reactor sites. Abandoned and disused mines, such as the salt mines in Germany, also

International perspective on repositories for low level waste

Energy Technology Data Exchange (ETDEWEB)

Bergstroem, Ulla; Pers, Karin; Almen, Ylva (SKB International AB (Sweden))

2011-12-15

Nuclear energy production gives rise to different types of radioactive waste. The use of nuclear isotopes within the research, industry and medical sectors also generates radioactive waste. To protect man and the environment from radiation the waste is isolated and contained by deposition in repositories. These repositories may have various designs regarding location, barriers etc depending on the potential danger of the waste. In Sweden, low- and intermediate level waste (LILW) is disposed of in the SFR repository in Forsmark. The repository is located 60 metres down into the bedrock under the bottom of the sea and covered by 6 metres of water. It is planned to extend SFR to accommodate decommissioning waste from the dismantling of the Swedish nuclear power facilities and also for the additional operation waste caused by the planned prolonged operation time. When planning the extension consultations will be carried out with the host municipality, authorities, organisations and general public. In planning the extension, SKB has performed a worldwide compilation of how other countries have, or plan to, handle the final disposal of similar wastes. The aim of this report is to give a brief description of LILW repositories worldwide; including general brief descriptions of many facilities, descriptions of the waste and the barriers as well as safety assessments for a few chosen repositories which represent different designs. The latter is performed, where possible, to compare certain features against the Swedish SFR. To provide a background and context to this study, international organisations and conventions are also presented along with internationally accepted principles regarding the management of radioactive waste. Similar to SFR, suitable locations for the repositories have, in many countries, been found at sites that already have, or used to have nuclear activities, such as reactor sites. Abandoned and disused mines, such as the salt mines in Germany, also

A Unique Digital Electrocardiographic Repository for the Development of Quantitative Electrocardiography and Cardiac Safety: The Telemetric and Holter ECG Warehouse (THEW)

Science.gov (United States)

Couderc, Jean-Philippe

2010-01-01

The sharing of scientific data reinforces open scientific inquiry; it encourages diversity of analysis and opinion while promoting new research and facilitating the education of next generations of scientists. In this article, we present an initiative for the development of a repository containing continuous electrocardiographic information and their associated clinical information. This information is shared with the worldwide scientific community in order to improve quantitative electrocardiology and cardiac safety. First, we present the objectives of the initiative and its mission. Then, we describe the resources available in this initiative following three components: data, expertise and tools. The Data available in the Telemetric and Holter ECG Warehouse (THEW) includes continuous ECG signals and associated clinical information. The initiative attracted various academic and private partners whom expertise covers a large list of research arenas related to quantitative electrocardiography; their contribution to the THEW promotes cross-fertilization of scientific knowledge, resources, and ideas that will advance the field of quantitative electrocardiography. Finally, the tools of the THEW include software and servers to access and review the data available in the repository. To conclude, the THEW is an initiative developed to benefit the scientific community and to advance the field of quantitative electrocardiography and cardiac safety. It is a new repository designed to complement the existing ones such as Physionet, the AHA-BIH Arrhythmia Database, and the CSE database. The THEW hosts unique datasets from clinical trials and drug safety studies that, so far, were not available to the worldwide scientific community. PMID:20863512

Establishing managerial requirements for low-and intermediate-level waste repository

International Nuclear Information System (INIS)

Chung, C. W.; Lee, Y. K.; Kim, H. T.; Park, W. J.; Suk, T. W.; Park, S. H.

2004-01-01

This paper reviews basic considerations for establishing managerial requirements on the domestic low-and intermediate-level radioactive waste repository and presents the corresponding draft requirements. The draft emphasizes their close linking with the related regulations, standards and safety assessment for the repository. It also proposes a desirable direction towards harmonizing together with the existing waste acceptance requirements for the repository

U.S. DEPARTMENT OF ENERGY EXPERIENCE IN CREATING AND COMMUNICATING THE CASE FOR THE SAFETY OF A POTENTIAL YUCCA MOUNTAIN REPOSITORY

International Nuclear Information System (INIS)

W.J. Boyle; A.E. Van Luik

2005-01-01

Experience gained by the U.S. Department of Energy (the Department) in making the recommendation for the development of the Yucca Mountain site as the nation's first high-level waste and spent nuclear fuel repository is useful for creating documents to support the next phase in the repository program, the licensing phase. The experience that supported the successful site-recommendation process involved a three-tiered approach. First, was making a highly technical case for regulatory compliance. Second, was making a broader case for safety in an Environmental Impact Statement. And third, producing plain language brochures, made available to the public in hard copy and on the Internet, to explain the Department's action and its legal and scientific bases. This paper reviews lessons learned from this process, and makes suggestions for the next stage of the repository program: licensing

Joint SKI and SSI review of SKB preliminary safety assessment of repository for long-lived low- and intermediate-level waste. Review report

International Nuclear Information System (INIS)

2001-03-01

SKI and SSI find that SKB's first proper safety assessment of the SFL 3-5 repositories provides a valuable springboard for continued efforts in this field. Even though the safety assessment is relatively limited in scope, it has numerous merits. The specific problems associated with the chosen repository concept for SFL 3-5 are discussed in a generally transparent manner. On the other hand, the authorities consider that SKB have only partly achieved the expressed goal of studying the significance of the current repository design and the choice of site. The greatest deficiency consists in that neither internal disturbances (such as considerable cracking or degradation of concrete structures) nor external disturbances (such as the effects of climate changes and glaciation) have been addressed in a thorough manner. A coherent report justifying the design choice from a long-term safety perspective is, in large part, not found here. SKI and SSI recommend that SKB provide a comparison with other possible SFL 3-5 repository designs. Depending upon, among other factors, what geospheric and biospheric conditions are assumed, SKB have shown that the calculated dose values could be relatively high for certain cases. More realistic assessments would be needed to draw reasonable comparisons between different sites, and to evaluate the importance of different nuclides in different contexts. Our review of SKBs preliminary safety assessment indicates that a great deal of research and development work remains to be done before the level of knowledge in this field is comparable with that associated with the final repository for spent fuel. This is reflected with unanimity in the international expert committee's review, and in the consultants' reviews. SKI and SSI wish to point out in particular the fact that comparison with SFR is of limited value, since the safety associated with SFL 3- 5 must be assessed on a much longer time scale. SKI and SSI find it remarkable that SKB have

A Study on Establishment of Buffer Zone of Radioactive Waste Repository

International Nuclear Information System (INIS)

Yoon, Jeong Hyoun; Park, Joo Wan; Ju, Min Su; Kim, Chang Lak; Park, Jin Baek

2008-01-01

A new proposed repository has a final capacity of 800,000 drums radioactive waste. Most of foreign repositories have a general practice of segregating control zones which mainly contributes to classification of degree of control, whether it is called buffer zone or not. Domestic regulatory requirements of establishment of buffer zone in a repository are not much different from those of nuclear power plants for operation period, in which satisfactory design objective or performance objective is the most important factor in determination of the buffer zone. The meaning of buffer zone after closure is a minimum requested area which can prevent inadvertent intruders from leading to non-allowable exposure during institutional control period. Safety assessment with drinking well scenario giving rise to the highest probability of exposure among the intruder's actions can verify fulfillment of the buffer zone which is determined by operational safety of the repository. At present. for the repository to be constructed in a few years, the same procedure and concept as described in this paper are applied that can satisfy regulatory requirements and radiological safety as well. However, the capacity of the repository will be stepwise extended upto 800,000 drums, consequently its layout will be varied too. Timely considerations will be necessary for current boundary of the buffer zone which has been established on the basis of 100,000 drums disposal.

Novi Han Radioactive Waste Repository post-closure safety assessment, ver.2

International Nuclear Information System (INIS)

Mateeva, M.

2003-01-01

The methodology for the post-closure safety assessment is presented. The assessment context includes regulatory framework (protection principles); scope and time frame; radiological and technical requirements; modeling etc. The description of the Novi Han disposal system contains site location. meteorological, hydrological and seismological characteristics; waste and repository description and human activities characteristics. The next step in the methodology is scenario development and justification. The systematic generation os exposure scenarios is considered as central to the post-closure safety assessment. The most important requirements for the systematic scenario generation approach are: transparency, comprehensiveness (all possible FEPs influencing the the disposal system and the radionuclide release should be considered); relevant future evolutions; identification of critical issues and investigation of the robustness of the system. For the source-pathway-receptor analysis the Process System is divided into near-field, geosphere/atmosphere and biosphere, describing the key facets controlling the potential radionuclide migration to the environment. The schematic division of the Novi Han near-field Process System into lower-level conceptual features is presented and discussed. As a result of the examinations of the FEPs three classes of scenarios are identified for the Novi Han post-closure safety assessment: Environmental evolution scenarios (geological change and climate change); future human action scenarios (human intrusion and archaeological action); Scenarios with very low probability (terrorism, crashes, explosions). The safety assessment iteration leads to identification of a modern scenario generation approach, assessment of key radionuclide releases, geological and hydrological evaluation, identification of the key parameters from sensitivity analysis etc. Examples of conceptual models are given. For the mathematical modeling the AMBER code is used

Earthquakes - a danger to deep-lying repositories?

International Nuclear Information System (INIS)

2012-03-01

This booklet issued by the Swiss National Cooperative for the Disposal of Radioactive Waste NAGRA takes a look at geological factors concerning earthquakes and the safety of deep-lying repositories for nuclear waste. The geological processes involved in the occurrence of earthquakes are briefly looked at and the definitions for magnitude and intensity of earthquakes are discussed. Examples of damage caused by earthquakes are given. The earthquake situation in Switzerland is looked at and the effects of earthquakes on sub-surface structures and deep-lying repositories are discussed. Finally, the ideas proposed for deep-lying geological repositories for nuclear wastes are discussed

Evaluation of Safety Culture Implementation and Socialization Results

International Nuclear Information System (INIS)

Situmorang, Johnny

2003-01-01

Evaluation of safety culture implementation and socialization results has been perform. Evaluation is carried out with specifying safety culture indicators, namely: Meeting between management and employee, system for incidents analysis, training activities related to improving safety, meeting with regulator, contractors, surveys on behavioural attitudes, and resources allocated to promote safety culture. Evaluation is based on observation and visiting the facilities to show the compliance indicator in term of good practices in the frame of safety culture implementation. For three facilities of research reactors, Kartini Yogyakarta, TRIGA Mark II Bandung and MPR-GAS Serpong, implementation of safety culture is considered good enough and progressive. Furthermore some indicator should be considered more intensive, for example the allocated resources, self assesment based on own questionnaire in the frame of improving the safety culture implementation. (author)

Preclosure radiological safety analysis for accident conditions of the potential Yucca Mountain Repository: Underground facilities

International Nuclear Information System (INIS)

Ma, C.W.; Sit, R.C.; Zavoshy, S.J.; Jardine, L.J.; Laub, T.W.

1992-06-01

This preliminary preclosure radiological safety analysis assesses the scenarios, probabilities, and potential radiological consequences associated with postulated accidents in the underground facility of the potential Yucca Mountain repository. The analysis follows a probabilistic-risk-assessment approach. Twenty-one event trees resulting in 129 accident scenarios are developed. Most of the scenarios have estimated annual probabilities ranging from 10 -11 /yr to 10 -5 /yr. The study identifies 33 scenarios that could result in offsite doses over 50 mrem and that have annual probabilities greater than 10 -9 /yr. The largest offsite dose is calculated to be 220 mrem, which is less than the 500 mrem value used to define items important to safety in 10 CFR 60. The study does not address an estimate of uncertainties, therefore conclusions or decisions made as a result of this report should be made with caution

Enlargement of the Baldone near-surface radioactive waste repository

International Nuclear Information System (INIS)

Dreimanis, A.

2007-01-01

A unified analysis of the enlargement of the Baldone near-surface radioactive waste (RW) repository RADONS considers the interplay of the existing engineering, safety and infrastructure premises, with the foreseen newly socio-technical features. This enlargement consists in construction of two additional RW disposal vaults and in building a long-term storage facility for spent sealed sources at the RADONS territory. Our approach is based on consecutive analysis of following basic elements: - the origin of enlargement - the RADONS safety analysis and a set of optimal socio-technical solutions of Salaspils research reactor decommissioning waste management; - the enlargement - a keystone of the national RW management concept, including the long-term approach; - the enlargement concept - the result of international co-operation and obligations; - arrangement optimization of new disposal and storage space; - environmental impact assessment for the repository enlargement - the update of socio-technical studies. The study of the public opinion revealed: negative attitude to repository enlargement is caused mainly due to missing information on radiation level and on the RADONS previous operations. These results indicate: basic measures to improve the public attitude to repository enlargement: the safety upgrade, public education and compensation mechanisms. A detailed stakeholders engagement and public education plan is elaborated. (author)

Long-term safety for KBS-3 repositories at Forsmark and Laxemar - a first evaluation. Main Report of the SR-Can project

International Nuclear Information System (INIS)

Hedin, Allan

2006-10-01

This document is the main report from the safety assessment project SR-Can. The SR-Can project is a preparatory stage for the SR-Site assessment, the report that will be used in support of SKB's application for a final repository. The purposes of the safety assessment SR-Can are the following: 1. To make a first assessment of the safety of potential KBS-3 repositories at Forsmark and Laxemar to dispose of canisters as specified in the application for the encapsulation plant. 2. To provide feedback to design development, to SKB's RandD programme, to further site investigations and to future safety assessment projects. 3. To foster a dialogue with the authorities that oversee SKB's activities, i.e. the Swedish Nuclear Power Inspectorate, SKI, and the Swedish Radiation Protection Authority, SSI, regarding interpretation of applicable regulations, as a preparation for the SR-Site project. The assessment relates to the KBS-3 disposal concept in which copper canisters with a cast iron insert containing spent nuclear fuel are surrounded by bentonite clay and deposited at approximately 500 m depth in saturated, granitic rock. Preliminary data from the Forsmark and Laxemar sites, presently being investigated by SKB as candidates for a KBS-3 repository are used in the assessment. An important aim of this report is to demonstrate the proper handling of requirements placed on the safety assessment in applicable regulations. Therefore, regulations issued by the Swedish Nuclear Power Inspectorate and the Swedish Radiation Protection Institute are reproduced in an Appendix where references are given to sections in the main text where the handling of the different requirements is discussed. The principal acceptance criterion requires that 'the annual risk of harmful effects after closure does not exceed 10 -6 for a representative individual in the group exposed to the greatest risk'. 'Harmful effects' refer to cancer and hereditary effects. The risk limit corresponds to an

Long-term safety for KBS-3 repositories at Forsmark and Laxemar - a first evaluation. Main Report of the SR-Can project

Energy Technology Data Exchange (ETDEWEB)

Hedin, Allan (ed.)

2006-10-15

This document is the main report from the safety assessment project SR-Can. The SR-Can project is a preparatory stage for the SR-Site assessment, the report that will be used in support of SKB's application for a final repository. The purposes of the safety assessment SR-Can are the following: 1. To make a first assessment of the safety of potential KBS-3 repositories at Forsmark and Laxemar to dispose of canisters as specified in the application for the encapsulation plant. 2. To provide feedback to design development, to SKB's RandD programme, to further site investigations and to future safety assessment projects. 3. To foster a dialogue with the authorities that oversee SKB's activities, i.e. the Swedish Nuclear Power Inspectorate, SKI, and the Swedish Radiation Protection Authority, SSI, regarding interpretation of applicable regulations, as a preparation for the SR-Site project. The assessment relates to the KBS-3 disposal concept in which copper canisters with a cast iron insert containing spent nuclear fuel are surrounded by bentonite clay and deposited at approximately 500 m depth in saturated, granitic rock. Preliminary data from the Forsmark and Laxemar sites, presently being investigated by SKB as candidates for a KBS-3 repository are used in the assessment. An important aim of this report is to demonstrate the proper handling of requirements placed on the safety assessment in applicable regulations. Therefore, regulations issued by the Swedish Nuclear Power Inspectorate and the Swedish Radiation Protection Institute are reproduced in an Appendix where references are given to sections in the main text where the handling of the different requirements is discussed. The principal acceptance criterion requires that 'the annual risk of harmful effects after closure does not exceed 10{sup -6} for a representative individual in the group exposed to the greatest risk'. 'Harmful effects' refer to cancer and hereditary effects

Geological boundary conditions for a safety demonstration and verification concept for a HLW repository in claystone in Germany. AnSichT

Energy Technology Data Exchange (ETDEWEB)

Stark, Lena; Bebiolka, Anke; Gerardi, Johannes [Federal Institute for Geosciences and Natural Resources (BGR), Hannover (Germany). Dept. of Underground Space for Storage and Economic Use; and others

2015-07-01

Within the framework of the R and D project ''AnSichT'', DBE TECHNOLOGY, BGR and GRS are developing a method to demonstrate the safety of a HLW repository in claystone in Germany. The methodological approach basing on a holistic concept, links the legal and geologic boundary conditions, the disposal and closure concept, the demonstration of barrier integrity, and the long-term analysis of the repository evolution as well. The geologic boundary conditions are specified by the description of the geological situation and generic models, the selection of representative parameters and geoscientific long-term predictions. They form a fundament for the system analysis.

Deep repository for spent nuclear fuel. SR-97-Post-closure safety. Main Report. Volume I and II

Energy Technology Data Exchange (ETDEWEB)

Hedin, A. [ed.

1999-11-01

In preparation for coming site investigations for siting of a deep repository for spent nuclear fuel, the Swedish Government and nuclear regulatory authorities have requested an assessment of the repository's long-term safety. The purpose is to demonstrate whether the risk of harmful effects in individuals in the vicinity of the repository complies with the acceptance criterion formulated by the Swedish regulatory authorities, i.e. that the risk may not exceed 10{sup -6} per year. Geological data are taken from three sites in Sweden to shed light on different conditions in Swedish granitic bedrock. The future evolution of the repository system is analyzed in the form of five scenarios. The first is a base scenario where the repository is postulated to be built entirely according to specifications and where present-day conditions in the surroundings are postulated to persist. The four other scenarios show how the evolution of the repository differs from that in the base scenario if the repository contains a few initially defective canisters, in the event of climate change, earthquakes, and future inadvertent human intrusion. The time horizon for the analyses is at most one million years, in accordance with preliminary regulations. By means of model studies and calculations, the base scenario analyzes how the radioactivity of the fuel declines with time, the repository's thermal evolution as a result of the decay heat in the fuel, the hydraulic evolution in buffer and backfill when they become saturated with water, and the long-term groundwater flow in the geosphere on the three sites. The overall conclusion of the analyses in the base scenario is that the copper canisters isolating capacity is not threatened by either the mechanical or chemical stresses to which it is subjected. The safety margins are great even in a million-year perspective. The internal evolution in initially defective canisters and the possible resultant migration of radionuclides in

Deep repository for spent nuclear fuel. SR-97-Post-closure safety. Main Report. Volume I and II

Energy Technology Data Exchange (ETDEWEB)

Hedin, A [ed.

1999-11-01

In preparation for coming site investigations for siting of a deep repository for spent nuclear fuel, the Swedish Government and nuclear regulatory authorities have requested an assessment of the repository's long-term safety. The purpose is to demonstrate whether the risk of harmful effects in individuals in the vicinity of the repository complies with the acceptance criterion formulated by the Swedish regulatory authorities, i.e. that the risk may not exceed 10{sup -6} per year. Geological data are taken from three sites in Sweden to shed light on different conditions in Swedish granitic bedrock. The future evolution of the repository system is analyzed in the form of five scenarios. The first is a base scenario where the repository is postulated to be built entirely according to specifications and where present-day conditions in the surroundings are postulated to persist. The four other scenarios show how the evolution of the repository differs from that in the base scenario if the repository contains a few initially defective canisters, in the event of climate change, earthquakes, and future inadvertent human intrusion. The time horizon for the analyses is at most one million years, in accordance with preliminary regulations. By means of model studies and calculations, the base scenario analyzes how the radioactivity of the fuel declines with time, the repository's thermal evolution as a result of the decay heat in the fuel, the hydraulic evolution in buffer and backfill when they become saturated with water, and the long-term groundwater flow in the geosphere on the three sites. The overall conclusion of the analyses in the base scenario is that the copper canisters isolating capacity is not threatened by either the mechanical or chemical stresses to which it is subjected. The safety margins are great even in a million-year perspective. The internal evolution in initially defective canisters and the possible resultant migration of radionuclides in buffer

Safety indicators in different time frames for the safety assessment of underground radioactive waste repositories. First report of the INWAC subgroup on principles and criteria for radioactive waste disposal

International Nuclear Information System (INIS)

1994-10-01

Principles and criteria for the disposal of long lived radioactive waste involve issues which go beyond those normally considered in the basic system of radiation protection. Safety criteria based on radiation risk an dose limitation are commonly accepted as the principal basis for judging the acceptability of radioactive waste repositories. However, the long time-scales of interest mean that risks or doses to future individuals cannot be predicted with any certainty as they depend, amongst other things, on assumptions made about the integrity of the waste matrix, the man-made barriers, the geology, the dispersion of groundwater, etc. and future biospheric conditions and human lifestyles. This document discusses various safety indicators and their applicability in the context of the future time-scales which have to be considered in safety assessments of deep geologic repositories. Quantitative assessment are based on numerical estimates of consequences (e.g. risk or dose) and the assessment is made against numerical criteria. Qualitative assessments are based on estimates of hazard potential which are not exact or absolute and the assessment is made against criteria which may not be numerically defined. Examples of such criteria are the convenient reference values provided by levels of radionuclides in the natural environment. Refs, figs and tabs

R and D on performance assessment of a potential LILW repository

Energy Technology Data Exchange (ETDEWEB)

Hwang, Yong Soo; Lee, Sung Hee; Han, Kyung Won; Lee, Y M

2001-01-01

In this technical report Important issues and assessment methods related to post-closure safety of a proposed repository has been discussed, regarding LILW disposal. At first, to summarize the new proposal on the amendment of the MOST Notice 96-16, KAERI set up the overall directions on future R and D over performance assessment. In addition relevant overseas project such as Finnish POSIVA's VLJ repository project was thoroughly reviewed along with the recent progress of the gas generation and migration R and D. In the post-closure safety analysis of the VLJ repository, in addition to the normal evolution scenario, several disturbed evolution and accident scenarios have been analysed. The groundwater flow analysis and the biosphere analysis have been evaluated. The result of the safety analysis show that radiation doses of any significance are caused only if a well is bored in the vicinity of the repository or if the groundwater discharge spot is inhabited and used for cultivation. in the reference scenario the maximum expectation value of the individual dose rate is 0.03 mSv/a. in the realistic scenario the maximum expectation value of the dose rate is 0.0002 mSv/a In general gas generation sources are represented by radioactive gases and non-radioactive gases. The amount of radioactive gases is little that it does not make a significant influence on the safety of LILW repository, however non-radioactive gases can cause safety problem. As mentioned above, gas of LILW generated by corrosion and microbiology and their production rate can be estimated by computer simulation and long-term experiment.

R and D on performance assessment of a potential LILW repository

Energy Technology Data Exchange (ETDEWEB)

Hwang, Yong Soo; Lee, Sung Hee; Han, Kyung Won; Lee, Y. M

2001-01-01

In this technical report Important issues and assessment methods related to post-closure safety of a proposed repository has been discussed, regarding LILW disposal. At first, to summarize the new proposal on the amendment of the MOST Notice 96-16, KAERI set up the overall directions on future R and D over performance assessment. In addition relevant overseas project such as Finnish POSIVA's VLJ repository project was thoroughly reviewed along with the recent progress of the gas generation and migration R and D. In the post-closure safety analysis of the VLJ repository, in addition to the normal evolution scenario, several disturbed evolution and accident scenarios have been analysed. The groundwater flow analysis and the biosphere analysis have been evaluated. The result of the safety analysis show that radiation doses of any significance are caused only if a well is bored in the vicinity of the repository or if the groundwater discharge spot is inhabited and used for cultivation. in the reference scenario the maximum expectation value of the individual dose rate is 0.03 mSv/a. in the realistic scenario the maximum expectation value of the dose rate is 0.0002 mSv/a In general gas generation sources are represented by radioactive gases and non-radioactive gases. The amount of radioactive gases is little that it does not make a significant influence on the safety of LILW repository, however non-radioactive gases can cause safety problem. As mentioned above, gas of LILW generated by corrosion and microbiology and their production rate can be estimated by computer simulation and long-term experiment.

Repository waste-handling operations, 1998

International Nuclear Information System (INIS)

Cottam, A.E.; Connell, L.

1986-04-01

The Civilian Radioactive Waste Management Program Mission Plan and the Generic Requirements for a Mined Geologic Disposal System state that beginning in 1998, commercial spent fuel not exceeding 70,000 metric tons of heavy metal, or a quantity of solidified high-level radioactive waste resulting from the reprocessing of such a quantity of spent fuel, will be shipped to a deep geologic repository for permanent storage. The development of a waste-handling system that can process 3000 metric tons of heavy metal annually will require the adoption of a fully automated approach. The safety and minimum exposure of personnel will be the prime goals of the repository waste handling system. A man-out-of-the-loop approach will be used in all operations including the receipt of spent fuel in shipping casks, the inspection and unloading of the spent fuel into automated hot-cell facilities, the disassembly of spent fuel assemblies, the consolidation of fuel rods, and the packaging of fuel rods into heavy-walled site-specific containers. These containers are designed to contain the radionuclides for up to 1000 years. The ability of a repository to handle more than 6000 pressurized water reactor spent-fuel rods per day on a production basis for approximately a 23-year period will require that a systems approach be adopted that combines space-age technology, robotics, and sophisticated automated computerized equipment. New advanced inspection techniques, maintenance by robots, and safety will be key factors in the design, construction, and licensing of a repository waste-handling facility for 1998

Investigation of the behaviour of a LILW superficial repository under aircraft impact

International Nuclear Information System (INIS)

Lo Frano, Rosa; Stefanini, Lorenzo

2016-01-01

Highlights: • Safety assessment of a LILW superficial repository. • Investigation of the consequences of an aircraft impact with fuel burning. • Experimental material properties. • Numerical simulation of aircraft impact with fuel burning accident by MSC.MARC"© code. • Demonstration that the overall integrity resulted is guaranteed. - Abstract: Safety and security are the two fundamental aspects to guarantee when designing a LILW superficial repository. Because of its safety concern, we have to prove, and build confidence in, the primary and secondary consequences of the crashing will be acceptable. These goals are obtained generally by means of safety assessment supported by calculations. This study is intended to investigate the performance of a superficial repository subjected to aircraft impact and fuel burning. To the purpose a superficial repository similar to that of El Cabril has been considered. Moreover to be confident the facility is safe and that the consequences of such a type of accident on the environment and humans are negligible, an appropriate safety assessment was carried out. The potential damage that aircraft impact could bring into the repository has been therefore analysed and discussed. To attain the intent load functions, calculated according to the Riera approach, and the maximum temperature reached by fuel during its combustion have been considered. FEM (thermo-mechanical) simulations have been done, by MSC"© Marc code, assuming damaging phenomena of concrete and material properties variation with the temperature. The obtained results showed that an empty superficial repository with a wall thickness, ranging from 0.7 to 1 m, is not sufficient to avoid penetration. Nevertheless even in presence of a reduced strength and of (cone) cracking and plugging, the overall integrity resulted guaranteed.

Investigation of the behaviour of a LILW superficial repository under aircraft impact

Energy Technology Data Exchange (ETDEWEB)

Lo Frano, Rosa, E-mail: rosa.lofrano@ing.unipi.it; Stefanini, Lorenzo

2016-04-15

Highlights: • Safety assessment of a LILW superficial repository. • Investigation of the consequences of an aircraft impact with fuel burning. • Experimental material properties. • Numerical simulation of aircraft impact with fuel burning accident by MSC.MARC{sup ©} code. • Demonstration that the overall integrity resulted is guaranteed. - Abstract: Safety and security are the two fundamental aspects to guarantee when designing a LILW superficial repository. Because of its safety concern, we have to prove, and build confidence in, the primary and secondary consequences of the crashing will be acceptable. These goals are obtained generally by means of safety assessment supported by calculations. This study is intended to investigate the performance of a superficial repository subjected to aircraft impact and fuel burning. To the purpose a superficial repository similar to that of El Cabril has been considered. Moreover to be confident the facility is safe and that the consequences of such a type of accident on the environment and humans are negligible, an appropriate safety assessment was carried out. The potential damage that aircraft impact could bring into the repository has been therefore analysed and discussed. To attain the intent load functions, calculated according to the Riera approach, and the maximum temperature reached by fuel during its combustion have been considered. FEM (thermo-mechanical) simulations have been done, by MSC{sup ©} Marc code, assuming damaging phenomena of concrete and material properties variation with the temperature. The obtained results showed that an empty superficial repository with a wall thickness, ranging from 0.7 to 1 m, is not sufficient to avoid penetration. Nevertheless even in presence of a reduced strength and of (cone) cracking and plugging, the overall integrity resulted guaranteed.

Application and further development of models for the final repository safety analyses on the clearance of radioactive materials for disposal. Final report; Anwendung und Weiterentwicklung von Modellen fuer Endlagersicherheitsanalysen auf die Freigabe radioaktiver Stoffe zur Deponierung. Abschlussbericht

Energy Technology Data Exchange (ETDEWEB)

Artmann, Andreas; Larue, Juergen; Seher, Holger; Weiss, Dietmar

2014-08-15

The project of application and further development of models for the final repository safety analyses on the clearance of radioactive materials for disposal is aimed to study the long-term safety using repository-specific simulation programs with respect to radiation exposure for different scenarios. It was supposed to investigate whether the 10 micro Sv criterion can be guaranteed under consideration of human intrusion scenarios. The report covers the following issues: selection and identification of models and codes and the definition of boundary conditions; applicability of conventional repository models for long-term safety analyses; modeling results for the pollutant release and transport and calculation of radiation exposure; determination of the radiation exposure.

Cross Institutional Cooperation on a Shared Bit Repository

DEFF Research Database (Denmark)

Zierau, Eld; Kejser, Ulla Bøgvad

2013-01-01

This paper explores how independent institutions, such as archives and libraries, can cooperate on managing a shared bit repository with bit preservation, in order to use their resources for preservation in a more cost-effective way. It uses the OAIS Reference Model to provide a framework...... for systematically analysing institutions technical and organisational requirements for a remote bit repository. Instead of viewing a bit repository simply as Archival Storage for the institutions repositories, we argue for viewing it as consisting of a subset of functions from all entities defined by the OAIS...... Reference Model. The work is motivated by and used in a current Danish feasibility study for establishing a national bit repository. The study revealed that depending on their missions and the collections they hold, the institutions have varying requirements e.g. for bit safety, accessibility...

Cross Institutional Cooperation on a Shared Bit Repository

DEFF Research Database (Denmark)

Zierau, Eld; Kejser, Ulla Bøgvad

2010-01-01

This paper explores how independent institutions, such as archives and libraries, can cooperate on managing a shared bit repository with bit preservation in order to use their resources for preservation n in a more cost-effective way. It uses the OAIS Reference Model to provide a framework...... for systematically analysing the technical and organizational requirements of institutions for a remote bit repository. Instead of viewing a bit repository simply as Archival Storage for the institutions’ repositories, we argue for viewing it as consisting of a subset of functions from all entities defined...... by the OAIS Reference Model. The work is motivated by and used in a current Danish feasibility study for establishing a national bit repository. The study revealed that depending on their missions and the collections they hold, the institutions have varying requirements, such as for bit safety, accessibility...

Evaluation of the damages in rocks caused by the construction of a repository

International Nuclear Information System (INIS)

Devillers, C.; Escalier des Orres, P.

1988-12-01

The Commissariat a l'Energie Atomique (French Atomic Energy Commission) has conducted a bibliographic study of the damages in the rock caused by the construction of a repository, and several hydraulic simulations, to appreciate the influence of these damages on the safety of the repository. These studies have led to the proposal of construction techniques in accordance safety requirements and industrial feasibility [fr

CONSIDERATIONS ON SOILS ISOLATIVE PROPERTIES FOR SITING OF A NEW NEAR-SURFACE RADIOACTIVE WASTE REPOSITORY IN POLAND IN THE LIGHT OF THE LONG TERM SAFETY

Directory of Open Access Journals (Sweden)

Monika Skrzeczkowska

2012-07-01

Full Text Available The paper presents a brief description of the occurrence of favorable isolative conditions for new surface radioactive waste repository in Poland. Selected soils may be used as a natural bottom layer or engineering barrier in multi-barrier system of RW repository. Currently, there is no regulation establishing standards for the bottom isolation, and the only quantifiable parameter with regard to water permeability is given for the repository objects, which in their case has to be lower than 10-9 m/s. For the purposes of this paper, treating on providing suitable bottom isolation for the new repository, this parameter has been transferred onto the consideration for soils suitability with a statement that it shall not be lower than the one given for the infrastructure. Submitted information should be taken into consideration by updating the information for the siting process according to IAEA Safety Standards.

Design premises for a KBS-3V repository based on results from the safety assessment SR-Can and some subsequent analyses

Energy Technology Data Exchange (ETDEWEB)

2009-11-15

The objective with this report is to: - provide design premises from a long term safety aspect of a KBS-3V repository for spent nuclear fuel, to form the basis for the development of the reference design of the repository. The design premises are used as input to the documents, called production reports, that present the reference design to be analysed in the long term safety assessment SR-Site. It is the aim that the production reports should verify that the chosen design complies with the design premises given in this report, whereas this report takes the burden of justifying why these design premises are relevant. The more specific aims and objectives with the production reports are provided in these reports. The following approach is used: - The reference design analysed in SR-Can is a starting point for setting safety related design premises for the next design step. - A few design basis cases, in accordance with the definition used in the regulation SSMFS 2008:211 and mainly related to the canister, can be derived from the results of the SR-Can assessment. From these it is possible to formulate some specific design premises for the canister. - The design basis cases involve several assumptions on the state of other barriers. These implied conditions are thus set as design premises for these barriers. - Even if there are few load cases on individual barriers that can be directly derived from the analyses, SR-Can provides substantial feedback on most aspects of the analysed reference design. This feedback is also formulated as design premises. - An important part of SR-Can Main report is the formulation and assessment of safety function indicator criteria. These criteria are a basis for formulating design premises, but they are not the same as the design premises discussed in the present report. Whereas the former should be upheld throughout the assessment period, the latter refer to the initial state and must be defined such that they give a margin for

Design premises for a KBS-3V repository based on results from the safety assessment SR-Can and some subsequent analyses

International Nuclear Information System (INIS)

2009-11-01

The objective with this report is to: - provide design premises from a long term safety aspect of a KBS-3V repository for spent nuclear fuel, to form the basis for the development of the reference design of the repository. The design premises are used as input to the documents, called production reports, that present the reference design to be analysed in the long term safety assessment SR-Site. It is the aim that the production reports should verify that the chosen design complies with the design premises given in this report, whereas this report takes the burden of justifying why these design premises are relevant. The more specific aims and objectives with the production reports are provided in these reports. The following approach is used: - The reference design analysed in SR-Can is a starting point for setting safety related design premises for the next design step. - A few design basis cases, in accordance with the definition used in the regulation SSMFS 2008:211 and mainly related to the canister, can be derived from the results of the SR-Can assessment. From these it is possible to formulate some specific design premises for the canister. - The design basis cases involve several assumptions on the state of other barriers. These implied conditions are thus set as design premises for these barriers. - Even if there are few load cases on individual barriers that can be directly derived from the analyses, SR-Can provides substantial feedback on most aspects of the analysed reference design. This feedback is also formulated as design premises. - An important part of SR-Can Main report is the formulation and assessment of safety function indicator criteria. These criteria are a basis for formulating design premises, but they are not the same as the design premises discussed in the present report. Whereas the former should be upheld throughout the assessment period, the latter refer to the initial state and must be defined such that they give a margin for

Reference repository design concept for bedded salt

Energy Technology Data Exchange (ETDEWEB)

Carpenter, D.W.; Martin, R.W.

1980-10-08

A reference design concept is presented for the subsurface portions of a nuclear waste repository in bedded salt. General geologic, geotechnical, hydrologic and geochemical data as well as descriptions of the physical systems are provided for use on generic analyses of the pre- and post-sealing performance of repositories in this geologic medium. The geology of bedded salt deposits and the regional and repository horizon stratigraphy are discussed. Structural features of salt beds including discontinuities and dissolution features are presented and their effect on repository performance is discussed. Seismic hazards and the potential effects of earthquakes on underground repositories are presented. The effect on structural stability and worker safety during construction from hydrocarbon and inorganic gases is described. Geohydrologic considerations including regional hydrology, repository scale hydrology and several hydrological failure modes are presented in detail as well as the hydrological considerations that effect repository design. Operational phase performance is discussed with respect to operations, ventilation system, shaft conveyances, waste handling and retrieval systems and receival rates of nuclear waste. Performance analysis of the post sealing period of a nuclear repository is discussed, and parameters to be used in such an analysis are presented along with regulatory constraints. Some judgements are made regarding hydrologic failure scenarios. Finally, the design and licensing process, consistent with the current licensing procedure is described in a format that can be easily understood.

Reference repository design concept for bedded salt

International Nuclear Information System (INIS)

Carpenter, D.W.; Martin, R.W.

1980-01-01

A reference design concept is presented for the subsurface portions of a nuclear waste repository in bedded salt. General geologic, geotechnical, hydrologic and geochemical data as well as descriptions of the physical systems are provided for use on generic analyses of the pre- and post-sealing performance of repositories in this geologic medium. The geology of bedded salt deposits and the regional and repository horizon stratigraphy are discussed. Structural features of salt beds including discontinuities and dissolution features are presented and their effect on repository performance is discussed. Seismic hazards and the potential effects of earthquakes on underground repositories are presented. The effect on structural stability and worker safety during construction from hydrocarbon and inorganic gases is described. Geohydrologic considerations including regional hydrology, repository scale hydrology and several hydrological failure modes are presented in detail as well as the hydrological considerations that effect repository design. Operational phase performance is discussed with respect to operations, ventilation system, shaft conveyances, waste handling and retrieval systems and receival rates of nuclear waste. Performance analysis of the post sealing period of a nuclear repository is discussed, and parameters to be used in such an analysis are presented along with regulatory constraints. Some judgements are made regarding hydrologic failure scenarios. Finally, the design and licensing process, consistent with the current licensing procedure is described in a format that can be easily understood

Review of Quality Assurance in SKB's Repository Research Experiments

International Nuclear Information System (INIS)

Hicks, T.W.

2007-01-01

SKB is preparing licence applications for a spent nuclear fuel encapsulation plant and repository which will be supported by the SR-Site safety report. A separate safety report, SR-Can, has been produced by SKB in preparation for the SR-Site report. SKI is in the process of reviewing the SR-Can safety report. In preparation for this review, and with a view to building confidence in SKB's research activities and understanding SKB's handling of data and other information, SKI has examined SKB's application of QA measures in the management and conduct of repository research and development projects that support the SR-Can safety assessment. These preliminary investigations will serve to support the preparation of more detailed quality and technical audits of SKB's repository safety assessment after the submission of a licence application. SKI's approach to this QA review is based on the consideration of quality-affecting aspects of a selection of SKB's research and development activities. As part of this review, SKI identified the need to examine quality-related aspects of some of the many experiments and investigations that form part of SKB's repository research programme. This report presents the findings of such a review, focusing on experiments concerned with the properties and performance of the engineered barrier system. First, in order to establish a broad understanding of QA requirements for repository scientific investigations, QA procedures implemented in the management of research and development activities for the low-level radioactive waste repository near Drigg in the UK and the Waste Isolation Pilot Plant and Yucca Mountain repository projects in the US were studied. The QA procedures for experiments and tests undertaken in these projects were compared with those implemented by SKB. Key findings are: QA programmes have been implemented for each repository development programme in response to regulatory requirements. The need for regular audits of the

Site selection for deep geologic repositories - Consequences for society, economy and environment

International Nuclear Information System (INIS)

2010-03-01

In a few years, Switzerland will make the decision regarding site selection for geological underground repositories for the storage of radioactive wastes. Besides the safety issue, many citizens are interested in how such a repository will affect environment, economy and society in the selected site's region. This brochure summarizes the results of many studies on the socio-economic impacts of nuclear waste repositories. Radioactive wastes must be stored in such a way that mankind and environment are safely protected for a long period of time. How this goal may be achieved, is already known: geologic deep repositories warrant long-term safety. For the oncoming years in Switzerland the question is where the repository will be built. The search for an appropriate site for a repository in the proposed regions will launch discussions. Within the participative framework the regions may bring their requests. The demonstration of the safety of potential repository sites has the highest priority in the selection process. In the third procedural step additional rock investigations will be made. The socio-economic studies and the experience with existing plants show that radioactive waste management plants can be built and operated in good agreement with environmental requirements. The radioactive wastes in a deep underground repository are stored many hundred meters below the Earth's surface. There, they are isolated from our vital space. Technical barriers and the surrounding dense rock confinement prevent the release of radioactive materials into the environment. A deep repository has positive consequences for the regional economy. It increases trade and value creation and creates work places. The socio-economic impacts practically extend over one century, but strongly vary with time; they are the largest during the building period. High life quality and a positive population development in the selected site region are compatible with a deep repository. A fair and

Post-closure radiation dose assessment for Yucca Mountain repository

International Nuclear Information System (INIS)

Jia Mingyan; Zhang Xiabin; Yang Chuncai

2006-01-01

A brief introduction of post-closure long-term radiation safety assessment results was represented for the yucca mountain high-level waste geographic disposal repository. In 1 million years after repository closure, for the higher temperature repository operating mode, the peak annual dose would be 150 millirem (120 millirem under the lower-temperature operating mode) to a reasonably maximally exposed individual approximately 18 kilometers (11 miles) from the repository. The analysis of a drilling intrusion event occurring at 30,000 years indicated a peak of the mean annual dose to the reasonably maximally exposed individual approximately 18 kilometers (11 miles) downstream of the repository would be 0.002 millirem. The analysis of an igneous activity scenario, including a volcanic eruption event and igneous intrusion event indicated a peak of the mean annual dose to the reasonably maximally exposed individual approximately 18 kilometers downstream of the repository would be 0.1 millirem. (authors)

VerSi. A method for the quantitative comparison of repository systems

Energy Technology Data Exchange (ETDEWEB)

Kaempfer, T.U.; Ruebel, A.; Resele, G. [AF-Consult Switzerland Ltd, Baden (Switzerland); Moenig, J. [GRS Braunschweig (Germany)

2015-07-01

Decision making and design processes for radioactive waste repositories are guided by safety goals that need to be achieved. In this context, the comparison of different disposal concepts can provide relevant support to better understand the performance of the repository systems. Such a task requires a method for a traceable comparison that is as objective as possible. We present a versatile method that allows for the comparison of different disposal concepts in potentially different host rocks. The condition for the method to work is that the repository systems are defined to a comparable level including designed repository structures, disposal concepts, and engineered and geological barriers which are all based on site-specific safety requirements. The method is primarily based on quantitative analyses and probabilistic model calculations regarding the long-term safety of the repository systems under consideration. The crucial evaluation criteria for the comparison are statistical key figures of indicators that characterize the radiotoxicity flux out of the so called containment-providing rock zone (einschlusswirksamer Gebirgsbereich). The key figures account for existing uncertainties with respect to the actual site properties, the safety relevant processes, and the potential future impact of external processes on the repository system, i.e., they include scenario-, process-, and parameter-uncertainties. The method (1) leads to an evaluation of the retention and containment capacity of the repository systems and its robustness with respect to existing uncertainties as well as to potential external influences; (2) specifies the procedures for the system analyses and the calculation of the statistical key figures as well as for the comparative interpretation of the key figures; and (3) also gives recommendations and sets benchmarks for the comparative assessment of the repository systems under consideration based on the key figures and additional qualitative

IAEA coordinated research project (CRP). The use of selected safety indicators (concentrations, fluxes) in the assessment of radioactive waste disposal. Report 8: Natural uranium fluxes and their use in repository safety assesment - implications for coupled model development

International Nuclear Information System (INIS)

Read, D.

2003-01-01

An international research project has been established by the IAEA to evaluate the potential of natural radioelement concentrations and fluxes as alternative safety indicators in the assessment of radioactive waste disposal. The Finnish regulatory agency STUK is contributing to this study in the form of a report series dealing specifically with geochemical and hydrological recycling of trace elements in the stable Fennoscandian shield environment. In parallel, STUK has embarked on the development of a coupled chemical transport model based on previous work undertaken in Russia. The two initiatives are closely linked and should provide STUK with many of the tools necessary for evaluating future Posiva submissions. In order to prepare for these activities, the information required for modelling needs to be specified and the application methodology refined. This report examines the major factors that will need to be considered. Implicit in the approach is the requirement for formal testing against data from geochemical systems, including uranium-thorium ore bodies and emissions from groundwater springs and thermal spas. Although many natural analogue projects have been undertaken over the years, few of these studies specifically formulated the link to safety assessment and not all are directly applicable to the Finnish case. Nevertheless, there is a considerable body of data available from Finland, much of which has yet to be processed. A comprehensive data set exists for Palmottu and is now being re-examined to ensure that it can be exploited to its full potential. The Cu-U deposit at Hyrkkoelae provides a useful adjunct. Quantifying the effects of a repository heat source emplaced in a region of low ambient geothermal gradient is difficult. Recourse will necessarily be made to investigations at thermal spas elsewhere in Europe but first the transferability of these data needs to be assessed. The usefulness of natural indicators extends well beyond comparison of

The Ec prototype repository project: implications of assessments for refining repository design

International Nuclear Information System (INIS)

Svemar, C.

2004-01-01

The most important issue in the evaluation of the repository performance is the long term safety of the repository. Analyses for this issue focuses on the 'steady state' conditions which start at the time when the repository has been saturated and the groundwater table returned to its normal level. The bentonite buffer around the canisters is saturated and homogeneous, and the canister is located exactly in the centre of the buffer. The backfill in the tunnel has been saturated as well and fills the earlier open spaces in the tunnel completely. The task of the activities taking places prior to the start of the 'steady state' conditions, like excavation, deposition, backfilling and sealing, with due consideration to the processes a consequences they may cause in the long run, is to provide for these 'ideal' conditions, as close as possible. While studying these activities in detail it has become obvious that development of methods and techniques needs to be carefully addressed before the decision is made on how to apply them in the repository. One general finding is that the situation in engineering of details is not that much different from the situation in geological characterisation of a site in detail; one more detail of engineering and the consequences it brings often complicates the situation rather than supports the solution prioritized so far. Many of the practical issues have been studied in the Prototype Repository project in the AEspoe Hard Rock Laboratory (Pusch et al., 2000). The Prototype Repository consists of two sections with four respectively two deposition holes with bentonite buffer and canister, the latter holding electrical heaters. The sections are separated by a concrete plug, and the whole test is to be separated from the rest of the laboratory by an outer plug. The project has two objectives: 1. To demonstrate the integrated function of tile deep repository components under realistic conditions and to compare results with models and

Ethical considerations surrounding nuclear waste repository siting and mitigation

International Nuclear Information System (INIS)

Peters, T.F.

1983-01-01

The potential long-term health and safety effects of the nuclear materials stored in repositories, the extremely long periods of time over which such materials may be dangerous, and the equity implications of the siting of a repository in any given area are unlike the issues involved in other large-scale projects. They involve major philosophical issues basic to human perspectives on social relationships and on insuring the future of mankind. Safety and permanence are the two basic criteria for determining whether a waste proposal is satisfactory. This chapter takes the approach of public (or micro) ethics, whose task is to 1) articulate and clarify public values relevant to a problem, 2) identify and evaluate public options, and 3) rank alternatives in some order of ethical preferability. It addresses the four major repository-related issues: uncertainty and risks, geographic equity, intergenerational ethics, and implementation ethics

Preclosure seismic design methodology for a geologic repository at Yucca Mountain. Topical report YMP/TR-003-NP

International Nuclear Information System (INIS)

1996-10-01

This topical report describes the methodology and criteria that the U.S. Department of Energy (DOE) proposes to use for preclosure seismic design of structures, systems, and components (SSCs) of the proposed geologic repository operations area that are important to safety. Title 10 of the Code of Federal Regulations, Part 60 (10 CFR 60), Disposal of High-Level Radioactive Wastes in Geologic Repositories, states that for a license to be issued for operation of a high-level waste repository, the U.S. Nuclear Regulatory Commission (NRC) must find that the facility will not constitute an unreasonable risk to the health and safety of the public. Section 60.131 (b)(1) requires that SSCs important to safety be designed so that natural phenomena and environmental conditions anticipated at the geologic repository operations area will not interfere with necessary safety functions. Among the natural phenomena specifically identified in the regulation as requiring safety consideration are the hazards of ground shaking and fault displacement due to earthquakes

Overview of adaptive phased management repository design development

International Nuclear Information System (INIS)

Russell, S.

2011-01-01

The Nuclear Waste Management Organization is implementing Adaptive Phased Management, Canada's plan for long-term management of used nuclear fuel. The organization is proceeding with the process for selecting a site in partnership with an informed and willing host community to safely and securely container and isolate used nuclear fuel in a deep geological repository in a suitable rock formation. Adaptive Phased Management is the culmination of more than 30 years of research, development and demonstration of repository concepts in Canada. Adaptive Phased Management uses a phased and adaptive step-wise approach to the multi-barrier system which is consistent with the long-term waste management approaches being developed in many other countries with nuclear power programs such as Sweden, Finland, Switzerland, the United Kingdom and France. The Nuclear Waste Management Organization is examining and developing conceptual designs for a deep geological repository and associated facilities for the placement of used nuclear fuel in long-lived containers. This paper will examine two of these generic conceptual designs which have recently been refined and updated. These conceptual designs will be used to support a pre-project review of repository design and safety by the Canadian Nuclear Safety Commission. (author)

NWTS program criteria for mined geologic disposal of nuclear waste: repository performance and development criteria. Public draft

Energy Technology Data Exchange (ETDEWEB)

None

1982-07-01

This document, DOE/NWTS-33(3) is one of a series of documents to establish the National Waste Terminal Storage (NWTS) program criteria for mined geologic disposal of high-level radioactive waste. For both repository performance and repository development it delineates the criteria for design performance, radiological safety, mining safety, long-term containment and isolation, operations, and decommissioning. The US Department of Energy will use these criteria to guide the development of repositories to assist in achieving performance and will reevaluate their use when the US Nuclear Regulatory Commission issues radioactive waste repository rules.

NWTS program criteria for mined geologic disposal of nuclear waste: repository performance and development criteria. Public draft

International Nuclear Information System (INIS)

1982-07-01

This document, DOE/NWTS-33(3) is one of a series of documents to establish the National Waste Terminal Storage (NWTS) program criteria for mined geologic disposal of high-level radioactive waste. For both repository performance and repository development it delineates the criteria for design performance, radiological safety, mining safety, long-term containment and isolation, operations, and decommissioning. The US Department of Energy will use these criteria to guide the development of repositories to assist in achieving performance and will reevaluate their use when the US Nuclear Regulatory Commission issues radioactive waste repository rules

Viability Assessment of a Repository at Yucca Mountain. Volume 2: Preliminary Design Concept for the Repository and Waste Package

Energy Technology Data Exchange (ETDEWEB)

None

1998-12-01

This volume describes the major design features of the Monitored Geologic Repository. This document is not intended to provide an exhaustive, detailed description of the repository design. Rather, this document summarizes the major systems and primary elements of the design that are radiologically significant, and references the specific technical documents and design analyses wherein the details can be found. Not all portions of the design are at the same level of completeness. Highest priority has been given to assigning resources to advance the design of the Monitored Geologic Repository features that are important to radiological safety and/or waste isolation and for which there is no NRC licensing precedent. Those features that are important to radiological safety and/or waste isolation, but for which there is an NRC precedent, receive second priority. Systems and features that have no impact on radiological safety or waste isolation receive the lowest priority. This prioritization process, referred to as binning, is discussed in more detail in Section 2.3. Not every subject discussed in this volume is given equal treatment with regard to the level of detail provided. For example, less detail is provided for the surface facility design than for the subsurface and waste package designs. This different level of detail is intentional. Greater detail is provided for those functions, structures, systems, and components that play key roles with regard to protecting radiological health and safety and that are not common to existing nuclear facilities already licensed by NRC. A number of radiological subjects are not addressed in the VA, (e.g., environmental qualification of equipment). Environmental qualification of equipment and other radiological safety considerations will be addressed in the LA. Non-radiological safety considerations such as silica dust control and other occupational safety considerations are considered equally important but are not addressed in

Performance assessment development for a LILW repository in Slovenia

International Nuclear Information System (INIS)

Zeleznik, N.; Mele, I.

2001-01-01

Simultaneously with the site selection process for a low and intermediate level radioactive waste (LILW) repository, the preliminary assessment of the influence of the specific disposal concept on the environment and on the population was developed. The performance assessment team, organized in 1997 by ARAO, prepared several basic studies in order to clarify the objectives of the performance and safety assessment (PA/SA) procedure. In 1999 also the first performance assessment of two safety cases (surface and underground) for generic site for a LILW repository was realized. In the year 2000 activities on PA/SA analyses continued. A systematic, generic list was prepared of all possible features, events and processes (FEP list) predictable for surface or underground LILW disposal in Slovenia. Recommended and selected were the most reliable scenarios with conceptual models for LILW disposal in normal and altered evolution conditions. New verification of the obtained results was done with more powerful and accurate models for the surface repository over an aquifer of lower water permeability and an underground repository in a plastic rock. The results for both generic cases under normal evolution scenarios showed that there is a negligible dose influence on members of the critical population due to the migration of radionuclides from the foreseen LILW repository. The results of the already performed work as well as plans for the future activities are presented in the paper.(author)

Ventilation System Strategy for a Prospective Korean Radioactive Waste Repository

International Nuclear Information System (INIS)

Kim, Jin; Kwon, Sang Ki

2005-01-01

In the stage of conceptual design for the construction and operation of the geologic repository for radioactive wastes, it is important to consider a repository ventilation system which serves the repository working environment, hygiene and safety of the public at large, and will allow safe maintenance like moisture content elimination in repository for the duration of the repositories life, construction/operation/closure, also allowing safe waste transportation and emplacement. This paper describes the possible ventilation system design criteria and requirements for the prospective Korean radioactive waste repositories with emphasis on the underground rock cavity disposal method in the both cases of low and medium-level and high-level wastes. It was found that the most important concept is separate ventilation systems for the construction (development) and waste emplacement (storage) activities. In addition, ventilation network system modeling, natural ventilation, ventilation monitoring systems and real time ventilation simulation, and fire simulation and emergency system in the repository are briefly discussed.

Some concepts of model uncertainty for performance assessments of nuclear waste repositories

International Nuclear Information System (INIS)

Eisenberg, N.A.; Sagar, B.; Wittmeyer, G.W.

1994-01-01

Models of the performance of nuclear waste repositories will be central to making regulatory decisions regarding the safety of such facilities. The conceptual model of repository performance is represented by mathematical relationships, which are usually implemented as one or more computer codes. A geologic system may allow many conceptual models, which are consistent with the observations. These conceptual models may or may not have the same mathematical representation. Experiences in modeling the performance of a waste repository representation. Experiences in modeling the performance of a waste repository (which is, in part, a geologic system), show that this non-uniqueness of conceptual models is a significant source of model uncertainty. At the same time, each conceptual model has its own set of parameters and usually, it is not be possible to completely separate model uncertainty from parameter uncertainty for the repository system. Issues related to the origin of model uncertainty, its relation to parameter uncertainty, and its incorporation in safety assessments are discussed from a broad regulatory perspective. An extended example in which these issues are explored numerically is also provided

Safeguarding of spent fuel conditioning and disposal in geological repositories

International Nuclear Information System (INIS)

Forsstroem, H.; Richter, B.

1997-01-01

Disposal of spent nuclear fuel in geological formations, without reprocessing, is being considered in a number of States. Before disposal the fuel will be encapsulated in a tight and corrosion resistant container. The method chosen for disposal and the design of the repository will be determined by the geological conditions and the very strict requirements on long-term safety. From a safeguards perspective spent fuel disposal is a new issue. As the spent fuel still contains important amounts of material under safeguards and as it can not be considered practicably irrecoverable in the repository, the IAEA has been advised not to terminate safeguards, even after closure of the repository. This raises a number of new issues where there could be a potential conflict of interests between safety and safeguards demands, in particular in connection with the safety principle that burdens on future generations should be avoided. In this paper some of these issues are discussed based on the experience gained in Germany and Sweden about the design and future operation of encapsulation and disposal facilities. The most important issues are connected to the required level of safeguards for a closed repository, the differences in time scales for waste management and safeguards, the need for verification of the fissile content in the containers and the possibility of retrieving the fuel disposed of. (author)

Safety first. Yes, but which safety?; Primat der Sicherheit. Ja, aber welche Sicherheit ist gemeint?

Energy Technology Data Exchange (ETDEWEB)

Roehlig, Klaus-Juergen [Technische Univ. Clausthal, Clausthal-Zellerfeld (Germany). Inst. fuer Endlagerforschung; Eckhardt, Anne [risicare GmbH, Zollikerberg (Switzerland)

2017-09-01

The site selection law in Germany and the final report of the final repository commission state the central objective to find a repository site that will guarantee safety for the next million of years. Decision makers, concerned and interested people have obviously different opinions and acceptance criteria with respect to the tools for the demonstration of safety (safety case). Possible solutions for a broad acceptance of safety definitions are discussed.

The post-closure radiological safety case for a spent fuel repository in Sweden - An international peer review of the SKB license-application study of March 2011

International Nuclear Information System (INIS)

2012-01-01

Sweden is at the forefront among countries developing plans for a deep geological repository of highly radioactive waste. There is no such repository in operation yet worldwide, but Sweden, Finland and France are approaching the licensing stage. At the request of the Swedish government, the NEA organised an international peer review of the post-closure radiological safety case produced by the Swedish Nuclear Fuel and Waste Management Company (SKB) in support of the application for a general licence to construct and operate a spent nuclear fuel geological repository in the municipality of Oesthammar. The purpose of the review was to help the Swedish government, the public and relevant organisations by providing an international reference regarding the maturity of SKB's spent fuel disposal programme vis-a-vis best practices in long-term disposal safety and radiological protection. The International Review Team (IRT) consisted of ten international specialists, who were free of conflict of interest with the SKB and brought complementary expertise to the review. This report provides the background and findings of the international peer review. The review's findings are presented at several levels of detail in order to be accessible to both specialist and non-specialist readers

Low- and intermediate-level waste repository-induced effects

Energy Technology Data Exchange (ETDEWEB)

Leupin, O.X.; Marschall, P.; Johnson, L.; Cloet, V.; Schneider, J. [National Cooperative for the Disposal of Radioactive Waste (NAGRA), Wettingen (Switzerland); Smith, P. [Safety Assessment Management Ltd, Henley-On-Thames, Oxfordshire (United Kingdom); Savage, D. [Savage Earth Associates Ltd, Bournemouth, Dorset (United Kingdom); Senger, R. [Intera Inc., Ennetbaden (Switzerland)

2016-10-15

This status report aims at describing and assessing the interactions of the radioactive waste emplaced in a low- and intermediate level waste (L/ILW) repository with the engineered materials and the Opalinus Clay host rock. The Opalinus Clay has a thickness of about 100 m in the proposed siting regions. Among other things the results are used to steer the RD and D programme of NAGRA. The repository-induced effects considered in this report are of the following broad types: - Thermal effects: i.e. effects arising principally from the heat generated by the waste and the setting of cement. - Rock-mechanical effects: i.e. effects arising from the mechanical disturbance to the rock caused by the excavation of the emplacement caverns and other underground structures. - Hydraulic and gas-related effects: i.e. the effects of repository resaturation and of gas generation, e.g. due to the corrosion of metals within the repository, on the host rock and engineered barriers. - Chemical effects: i.e. chemical interactions between the waste, the engineered materials and the host rock. Deep geological repositories are designed to avoid or mitigate the impact of potentially detrimental repository-induced effects on long-term safety. For the repository under consideration in the present report, an assessment of those repository-induced effects that remain shows that detrimental chemical and mechanical impacts are largely confined to the rock adjacent to the excavations, thermal impacts are minimal and gas effects can be mitigated by appropriate design measures to reduce gas production and provide pathways for gas transport that limit gas pressure build-up (engineered gas transport system, or EGTS). Specific measures that are part of the current reference design are discussed in relation to their significance with respect to repository-induced effects. The disposal system described in this report provides a system of passive barriers with multiple safety functions. The disposal

Low- and intermediate-level waste repository-induced effects

International Nuclear Information System (INIS)

Leupin, O.X.; Marschall, P.; Johnson, L.; Cloet, V.; Schneider, J.; Smith, P.; Savage, D.; Senger, R.

2016-10-01

This status report aims at describing and assessing the interactions of the radioactive waste emplaced in a low- and intermediate level waste (L/ILW) repository with the engineered materials and the Opalinus Clay host rock. The Opalinus Clay has a thickness of about 100 m in the proposed siting regions. Among other things the results are used to steer the RD and D programme of NAGRA. The repository-induced effects considered in this report are of the following broad types: - Thermal effects: i.e. effects arising principally from the heat generated by the waste and the setting of cement. - Rock-mechanical effects: i.e. effects arising from the mechanical disturbance to the rock caused by the excavation of the emplacement caverns and other underground structures. - Hydraulic and gas-related effects: i.e. the effects of repository resaturation and of gas generation, e.g. due to the corrosion of metals within the repository, on the host rock and engineered barriers. - Chemical effects: i.e. chemical interactions between the waste, the engineered materials and the host rock. Deep geological repositories are designed to avoid or mitigate the impact of potentially detrimental repository-induced effects on long-term safety. For the repository under consideration in the present report, an assessment of those repository-induced effects that remain shows that detrimental chemical and mechanical impacts are largely confined to the rock adjacent to the excavations, thermal impacts are minimal and gas effects can be mitigated by appropriate design measures to reduce gas production and provide pathways for gas transport that limit gas pressure build-up (engineered gas transport system, or EGTS). Specific measures that are part of the current reference design are discussed in relation to their significance with respect to repository-induced effects. The disposal system described in this report provides a system of passive barriers with multiple safety functions. The disposal

Public concerns and choices regarding nuclear-waste repositories

International Nuclear Information System (INIS)

Rankin, W.L.; Nealey, S.M.

1981-06-01

Survey research on nuclear power issues conducted in the late 1970's has determined that nuclear waste management is now considered to be one of the most important nuclear power issues both by the US public and by key leadership groups. The purpose of this research was to determine the importance placed on specific issues associated with high-level waste disposal. In addition, policy option choices were asked regarding the siting of both low-level and high-level nuclear waste repositories. A purposive sampling strategy was used to select six groups of respondents. Averaged across the six respondent groups, the leakage of liquid wastes from storage tanks was seen as the most important high-level waste issue. There was also general agreement that the issue regarding water entering the final repository and carrying radioactive wastes away was second in importance. Overall, the third most important issue was the corrosion of the metal containers used in the high-level waste repository. There was general agreement among groups that the fourth most important issue was reducing safety to cut costs. The fifth most important issue was radioactive waste transportation accidents. Overall, the issues ranked sixth and seventh were, respectively, workers' safety and earthquakes damaging the repository and releasing radioactivity. The eighth most important issue, overall, was regarding explosions in the repository from too much radioactivity, which is something that is not possible. There was general agreement across all six respondent groups that the two least important issues involved people accidentally digging into the site and the issue that the repository might cost too much and would therefore raise electricity bills. These data indicate that the concerns of nuclear waste technologists and other public groups do not always overlap

US Department of Energy Approach to Probabilistic Evaluation of Long-Term Safety for a Potential Yucca Mountain Repository

International Nuclear Information System (INIS)

Dr. R. Dyer; Dr. R. Andrews; Dr. A. Van Luik

2005-01-01

Regulatory requirements being addressed in the US geological repository program for spent nuclear fuel and high-level waste disposal specify probabilistically defined mean-value dose limits. These dose limits reflect acceptable levels of risk. The probabilistic approach mandated by regulation calculates a ''risk of a dose,'' a risk of a potential given dose value at a specific time in the future to a hypothetical person. The mean value of the time-dependent performance measure needs to remain below an acceptable level defined by regulation. Because there are uncertain parameters that are important to system performance, the regulation mandates an analysis focused on the mean value of the performance measure, but that also explores the ''full range of defensible and reasonable parameter distributions''...System performance evaluations should not be unduly influenced by...''extreme physical situations and parameter values''. Challenges in this approach lie in defending the scientific basis for the models selected, and the data and distributions sampled. A significant challenge lies in showing that uncertainties are properly identified and evaluated. A single-value parameter has no uncertainty, and where used such values need to be supported by scientific information showing the selected value is appropriate. Uncertainties are inherent in data, but are also introduced by creating parameter distributions from data sets, selecting models from among alternative models, abstracting models for use in probabilistic analysis, and in selecting the range of initiating event probabilities for unlikely events. The goal of the assessment currently in progress is to evaluate the level of risk inherent in moving ahead to the next phase of repository development: construction. During the construction phase, more will be learned to inform a new long-term risk evaluation to support moving to the next phase: accepting waste. Therefore, though there was sufficient confidence of safety

Surface facilities for geological deep repositories - Measures against dangers during construction and operation

International Nuclear Information System (INIS)

2013-09-01

This brochure published by the Swiss National Cooperative for the Disposal of Radioactive Waste (NAGRA) discusses the measures that are to be taken to address the dangers encountered during the construction and operation of deep geological repositories for nuclear wastes. Firstly, the operation of such repositories during the emplacement of nuclear wastes is discussed and examples of possible repositories for fuel rods and highly-radioactive waste are presented. Various emission-protection issues and safety measures to be taken during construction of such repositories are looked at as is the protection of ground water. Safety considerations during the operational phase are discussed, including inclusion methods used for the wastes and radiation protection. The handling of radioactive wastes, the recognition of dangers and measures to be taken to counteract them are discussed. Various possible accidents are looked at

The project ANSICHT. Safety and demonstration methodology for a final repository in clay formations in Germany; Projekt ANSICHT. Sicherheits- und Nachweismethodik fuer ein Endlager im Tongestein in Deutschland. Synthesebericht

Energy Technology Data Exchange (ETDEWEB)

Jobmann, Michael; Bebiolka, Anke; Jahn, Steffen; and others

2017-03-30

Based on the status of science and technology and under consideration of international repository concepts the fundamental methodology for safety demonstration for a high-level radioactive waste final repository in clay formations Germany was developed. Basic elements of the safety concept are the geological site description and the geo-scientific long-term prognosis on future performance. Another important section is the closure and sealing concept for the mine shafts. In the frame of the project the fundamental elements were developed and documented for model regions in northern and southern Germany. Three independent safety proofs have to be performed: the demonstration of the geological barrier integrity (clay), the demonstration of the geo-technical barrier system integrity - i.e. closure constructions and backfilling of the shafts, and the radiological demonstration that the radionuclide release in the area is lower than the respective limiting value.

Surface-type repository for low and intermediate level radioactive waste in the Republic of Croatia

International Nuclear Information System (INIS)

Kucar-Dragicevic, S.; Zarkovic, V.; Subasic, D.

1995-01-01

The low-level intermediate-level (LL/IL) radioactive waste repository siting and construction project is one of the activities related to establishing the rad waste management system in the Republic of Croatia. The repository project design is one in an array of project activities which also include the site selection procedure and public attitude issues. The prepared design documentation gives technical, safety and financial background relevant for making a final decision on the waste disposal type, and it includes the technological, mechanical, civil and financial documentation on the preliminary/basic design level. During the last few years, the preliminary design has been prepared and safety assessment conducted for the tunnel-type LL/IL rad waste repository. As the surface-type repository is one of alternatives for final disposal the design documentation for that repository type was prepared during 1994. (author)

Biosphere models for safety assesment of radioactive waste disposal

Energy Technology Data Exchange (ETDEWEB)

Proehl, G; Olyslaegers, G; Zeevaert, T [SCK/CEN, Mol (Belgium); Kanyar, B [University of Veszprem (Hungary). Dept. of Radiochemistry; Pinedo, P; Simon, I [Centro de Investigaciones Energeticas Medioambientales y Tecnologicas (CIEMAT), Madrid (Spain); Bergstroem, U; Hallberg, B [Studsvik Ecosafe, Nykoeping (Sweden); Mobbs, S; Chen, Q; Kowe, R [NRPB, Chilton, Didcot (United Kingdom)

2004-07-01

The aim of the BioMoSA project has been to contribute in the confidence building of biosphere models, for application in performance assessments of radioactive waste disposal. The detailed objectives of this project are: development and test of practical biosphere models for application in long-term safety studies of radioactive waste disposal to different European locations, identification of features, events and processes that need to be modelled on a site-specific rather than on a generic base, comparison of the results and quantification of the variability of site-specific models developed according to the reference biosphere methodology, development of a generic biosphere tool for application in long term safety studies, comparison of results from site-specific models to those from generic one, Identification of possibilities and limitations for the application of the generic biosphere model. (orig.)

The Morsleben radwaste repository. Preparing for decommissioning

International Nuclear Information System (INIS)

Mehnert, M.; Schmitt, R.

2001-06-01

The publication is intended to illustrate with a brief chronology the history and the present situation of the Morsleben radwaste repository, including specific aspects such as the geology of the site and construction and engineering activities, the particulars of waste form emplacement and log-term storage conditions, topical issues relating to radiological safety during operation and after decommissioning. The brochure is designed for the general audience interested in background information on all aspects of the uses, operation and decommissioning of a radwaste repository in Germany. (orig./CB) [de

Production, consumption and transport of gases in deep geological repositories according to the Swiss disposal concept

International Nuclear Information System (INIS)

Diomidis, N; Cloet, V.; Leupin, O.X.; Marschall, P.; Poller, A.; Stein, M.

2016-12-01

In a deep geological repository for radioactive waste, in absence of oxygen and in presence of water, corrosion of various metals and alloys will lead to the formation of hydrogen. If present, organic materials may slowly degrade and generate carbon dioxide, methane and other gaseous species. Depending on local conditions, gaseous species can be consumed by chemical reactions and by microbial activity. If the resulting rate of gas generation exceeds the rate of migration of dissolved gas molecules in the pores of the engineered barriers or the host rock, the solubility limit of the gas will eventually be exceeded and the formation of a discrete gas phase will occur. Gases could continue to accumulate until the pressure becomes sufficient to be released in gaseous form. This report deals with the evolution of gas-related processes that can influence the long-term behaviour and safety of low- and intermediate-level waste (L/ILW) and high-level waste (HLW) repositories in Opalinus Clay. The main aim is to present a synthesis of processes and phenomena related to repository-produced gases and to assess their influence on repository performance. A current overview of gas sources, reactions and interactions, generation, consumption, and transport is provided. Furthermore, current scientific understanding is used to define safety function indicators and criteria, which are employed to evaluate the potential influence of repository-generated gas on safety-relevant properties of engineered and natural barriers. The assessment of gas generation, consumption and transport is addressed separately for the HLW and the L/ILW deep geological repositories. The employed methodology, which is common for both repository types, consists of the description and quantification of the potential gas sources, which include the waste, barrier components such as disposal canisters and other gas-generating repository components, and of the processes and reactions leading to the generation or

Production, consumption and transport of gases in deep geological repositories according to the Swiss disposal concept

Energy Technology Data Exchange (ETDEWEB)

Diomidis, N; Cloet, V.; Leupin, O.X.; Marschall, P.; Poller, A.; Stein, M.

2016-12-15

In a deep geological repository for radioactive waste, in absence of oxygen and in presence of water, corrosion of various metals and alloys will lead to the formation of hydrogen. If present, organic materials may slowly degrade and generate carbon dioxide, methane and other gaseous species. Depending on local conditions, gaseous species can be consumed by chemical reactions and by microbial activity. If the resulting rate of gas generation exceeds the rate of migration of dissolved gas molecules in the pores of the engineered barriers or the host rock, the solubility limit of the gas will eventually be exceeded and the formation of a discrete gas phase will occur. Gases could continue to accumulate until the pressure becomes sufficient to be released in gaseous form. This report deals with the evolution of gas-related processes that can influence the long-term behaviour and safety of low- and intermediate-level waste (L/ILW) and high-level waste (HLW) repositories in Opalinus Clay. The main aim is to present a synthesis of processes and phenomena related to repository-produced gases and to assess their influence on repository performance. A current overview of gas sources, reactions and interactions, generation, consumption, and transport is provided. Furthermore, current scientific understanding is used to define safety function indicators and criteria, which are employed to evaluate the potential influence of repository-generated gas on safety-relevant properties of engineered and natural barriers. The assessment of gas generation, consumption and transport is addressed separately for the HLW and the L/ILW deep geological repositories. The employed methodology, which is common for both repository types, consists of the description and quantification of the potential gas sources, which include the waste, barrier components such as disposal canisters and other gas-generating repository components, and of the processes and reactions leading to the generation or

Safety-relevant assessment concerning the proposals for investigations in proposed locations for stage 3 of the deep geological repositories project; Sicherheitstechnisches Gutachten zum Vorschlag der in Etappe 3 SGT weiter zu untersuchenden geologischen Standortgebiete - Sachplan geologische Tiefenlager, Etappe 2

Energy Technology Data Exchange (ETDEWEB)

NONE

2017-04-15

This comprehensive report published by the Swiss Federal Nuclear Safety Inspectorate ENSI examines the proposals made by the Swiss National Cooperative for the disposal of Radioactive Waste (NAGRA) concerning proposed locations for deep waste repositories. Important basic considerations to be made when narrowing down the choice of host rock types and the selection of possible locations are discussed. These include, amongst others, safety concepts for the repositories, geological models and technical feasibility of repositories in Opalinus Clay, Brown Dogger layers and marl. The properties of various host rocks are discussed. Methods to be used in choosing at least two possible sites for repositories are considered. Also, the radioactive dosage to be expected at the various sites proposed is discussed. A safety-relevant comparison of the sites is made and an overall assessment is presented. The locations proposed for surface facilities at three proposed locations are discussed. The report is completed with an appendix containing tables and a list of relevant expert reports.

Source terms for analysis of accidents at a high level waste repository

International Nuclear Information System (INIS)

Mubayi, V.; Davis, R.E.; Youngblood, R.

1989-01-01

This paper describes an approach to identifying source terms from possible accidents during the preclosure phase of a high-level nuclear waste repository. A review of the literature on repository safety analyses indicated that source term estimation is in a preliminary stage, largely based on judgement-based scoping analyses. The approach developed here was to partition the accident space into domains defined by certain threshold values of temperature and impact energy density which may arise in potential accidents and specify release fractions of various radionuclides, present in the waste form, in each domain. Along with a more quantitative understanding of accident phenomenology, this approach should help in achieving a clearer perspective on scenarios important to preclosure safety assessments of geologic repositories. 18 refs., 3 tabs

Memory provisions for the Manche Surface Repository

International Nuclear Information System (INIS)

Dumont, Jean-Noel; Espiet-Subert, Florence

2015-01-01

The French La Manche repository site received its last radioactive waste package in 1994. In 2003, the official surveillance phase of the closed repository started under the supervision of Andra (the national industrial operator), the French Nuclear Safety Authority (ASN) and society (e.g. the local municipalities). Florence Espiet explained that information on the existence of the repository, its content, how it was operated and how it works needs to be preserved. It also is planned to review the information periodically for a minimum of 300 years. She described the creation of two documents on memory (a detailed and a summary one), both on permanent paper, and the preservation of the land registration. The latter constitutes 'passive' provisions for preserving memory. In addition, a number of 'active' provisions are and will be put in place: guided visits, exhibitions, partnerships with organisations dealing with memory preservation, and the creation of a think tank. The latter consists of local citizens and politicians, retired employees from Andra and artists that meet several times a year and reflect on memory preservation from the perspective of, for instance, local history, education, arts and rituals. Finally, two types of markers will be used to preserve the repository's memory: i) three herbaria cataloguing the plants growing on the site of the repository, including a very short description of the repository, will be stored at different sites in France; ii) a stele indicating the main characteristics of the repository, potentially linked to an art work, will be erected at the repository

Performance analysis of a repository for low and intermediate level reactor waste

International Nuclear Information System (INIS)

Vieno, T.; Nordman, H.; Vuori, S.; Peltonen, E.

1987-01-01

In Finland, utilities producing nuclear energy are responsible for the management of the radioactive waste, including final disposal. As regards low and intermediate level waste, the approach has been adopted to employ the power plant sites for locations of repositories. The repositories will be excavated at the depth of about 50 to 125 m in the bedrock of the two Finnish nuclear power plant sites, Loviisa and Olkiluoto. The performance analysis presented in this paper has been carried out for the Preliminary Safety Analysis Report (PSAR) of the Olkiluoto repository. A flexible model has been developed to estimate the release of radionuclides from waste packages and their subsequent transport through the engineered barriers in the repository. Gradual degradation of the engineered barriers is accounted for by altering parameters at fixed time points. Safety margins of the disposal concept have been evaluated by including disturbed evolution scenarios in the analysis. 13 references, 10 figures, 2 tables

Role of waste packages in the safety of a high level waste repository in a deep geological formation

International Nuclear Information System (INIS)

Bretheau, F.; Lewi, J.

1990-06-01

The safety of a radioactive waste disposal facility lays on the three following barriers placed between the radioactive materials and the biosphere: the waste package; the engineered barriers; the geological barrier. The function assigned to each of these barriers in the performance assessment is an option taken by the organization responsible for waste disposal management (ANDRA in France), which must show that: expected performances of each barrier (confinement ability, life-time, etc.) are at least equal to those required to fulfill the assigned function; radiation protection requirements are met in all situations considered as credible, whether they be the normal situation or random event situations. The French waste management strategy is based upon two types of disposal depending on the nature and activity of waste packages: - surface disposal intended for low and medium level wastes having half-lives of about 30 years or less and alpha activity less than 3.7 MBq/kg (0.1 Ci/t), for individual packages and less than 0.37 MBq/kg (0.01 Ci/t) in the average. Deep geological disposal intended for TRU and high level wastes. The conditions of acceptance of packages in a surface disposal site are subject to the two fundamental safety rules no. I.2 and III.2.e. The present paper is only dealing with deep geological disposal. For deep geological repositories, three stages are involved: stage preceding definitive disposal (intermediate storage, transportation, handling, setting up in the disposal cavities); stage subsequent to definitive sealing of the disposal cavities but prior to the end of operation of the repository; stage subsequent to closure of the repository. The role of the geological barrier has been determined as the essential part of long term radioactivity confinement, by a working group, set up by the French safety authorities. Essential technical criteria relating to the choice of a site so defined by this group, are the following: very low permeability

Groundwater flow modeling of periods with temperate climate conditions for use in a safety assessment of a repository for spent nuclear fuel - 59154

International Nuclear Information System (INIS)

Joyce, Steven; Hartley, Lee; Simpson, Trevor

2012-01-01

Document available in abstract form only. Full text of publication follows: As a part of the license application for a final repository for spent nuclear fuel, the Swedish Nuclear Fuel and Waste Management Company (SKB) has prepared a safety report (SR-Site) that assesses the long-term radiological safety after closure of a repository located at 500 m depth in the Forsmark area, c. 120 km north of Stockholm. The movement and composition of groundwater affect both the key pathways for radionuclide migration and the performance of engineered barriers, and hence are important issues that have to be considered and modelled as part of quantitative assessment calculations. This presentation describes the groundwater flow modelling studies that have been performed to represent the post-closure hydrogeological and hydrochemical situations during temperate climate conditions, and how these are used to support safety assessment calculations and arguments. The collation and implementation of onsite hydrogeological and hydrogeochemical data from the surface based site investigations at Forsmark are used as the basis for defining a reference case for the natural hydrogeological situation at the site (hydrogeological base case). Areas of uncertainty within the current site understanding and the engineered system are examined by a series of flow model variants

Environmental monitoring and radiation protection programs of Novi Han radioactive waste repository

International Nuclear Information System (INIS)

Christoskova, M.; Kostova, M.; Sheherov, L.; Bekiarov, P.; Iovtchev, M.

2000-01-01

The system for monitoring and control as an important part of the safety management of the Novi Han Radioactive Waste Repository contains two independent programs: environmental monitoring of the site (controlled area), the restricted access area and the surveillance area (supervised area) of the repository and radiation protection program including personal dosimetric control and indoor dosimetric control of workplaces in the buildings of the repository. The main activities related to the programs implementation are presented

Classifying decommissioning wastes for allocation to appropriate final repositories

International Nuclear Information System (INIS)

Alder, J.C.; Tunaboylu, K.

1982-01-01

For the safe disposal of radioactive wastes in different repositories, it is of advantage to classify them in well-defined conditioned categories, appropriate for final disposal. These categories, the so-called waste sorts are characterized by similar radionuclide distribution, similar nuclide-specific activity concentrations and similar waste matrix. A methodology is presented for classifying decommissioning wastes and is applied to the decommissioning wastes arising from a Swiss program of 6 GWe. The amounts and nuclide-specific activity inventories of the decommissioning waste sorts have been estimated. A first allocation into two different repository types has been performed. Such a classification enables one to define the source parameters for repository safety analysis and allows one to allocate the different waste categories into appropriate final repositories. This work presents a first iteration to determine which waste sorts belong to which repository type. The characteristics of waste sorts have to be better defined and the protective strength of the repository barriers has to be optimized. 7 references, 2 figures, 4 tables

SR 97: post-closure safety of a deep repository for spent nuclear fuel in Sweden

International Nuclear Information System (INIS)

2000-01-01

A major activity of the Nuclear Energy Agency (NEA) in the field of radioactive waste management is the organisation of independent, international peer reviews of national studies and projects. The NEA peer reviews help national programmes to assess their achievements. The review reports also provide reference information to be shared with others on what is desirable and what is feasible. This report presents the common views of the International Review Team (IRT) established by the NEA Secretariat on behalf of the Swedish Nuclear Power Inspectorate (SKI) to perform a peer review of a post-closure safety study of a deep repository for spent nuclear fuel in Sweden, Safety Report 97, produced by the Swedish Spent Fuel and Waste Management Company (SKB). The review is based on the main reports of the project and supporting documents, on information exchanged with SKB staff both through the intermediary of SKI and in direct interaction at a week-long workshop in Sweden, on a visit of the SKB's Aespoe Hard Rock Laboratory and Canister Laboratory, as well as on internal discussions within the IRT. (authors)

General construction requirements for the deep repository in the KBS-3 system

International Nuclear Information System (INIS)

2002-10-01

The KBS-3 systems includes equipment and plants for transport of spent nuclear fuels and encapsulated spent fuels, central intermediate storage, encapsulation and deep geologic disposal. The requirements in this document concern the repository and have been put together in view of the tasks of designing, constructing and building the repository. The report presents: A general review of existing design plans; Laws and regulations relevant for the design of the repository; How the regulations have been broken down to functional demands and dimensioning requirements for the repository; How the site conditions influence the design, and how the layout of the different parts of the repository interact; Relations between the functions of the repository, the safety and the design; A foundation for developing construction plans for the repository. The requirements will be collected in a database that will develop as new knowledge is collected

The main demands and criteria for building site choice for radioactive waste repositories

International Nuclear Information System (INIS)

Angelova, R.; Sandul, G.A.; Sen'ko, T.Ya.

2002-01-01

There are considered the main demands of building site choice for RAW repositories. At this the accent is placed on geological repositories (underground repositories of geological type) and near surface repositories assigned to disposal of low- and intermediate-level short- and mediate-lived radionuclides. These demands are conditionally separated into two blocks: account of social development of the adjoining territories; account of natural factors characterizing building site. Further there are discussed the questions of anthropogenous influence on a safety functioning of RAW repositories and of urgency of stable development of the adjoining territories. In context of the Ukrainian and other states nuclear laws there is also considered the lawful aspect defining the building site choice for RAW repositories

Long-term safety assessment of trench-type surface repository at Chernobyl, Ukraine - computer model and comparison with results from simplified models

International Nuclear Information System (INIS)

Haverkamp, B.; Krone, J.; Shybetskyi, I.

2013-01-01

The Radioactive Waste Disposal Facility (RWDF) Buryakovka was constructed in 1986 as part of the intervention measures after the accident at Chernobyl NPP (ChNPP). Today, the surface repository for solid low and intermediate level waste (LILW) is still being operated but its maximum capacity is nearly reached. Long-existing plans for increasing the capacity of the facility shall be implemented in the framework of the European Commission INSC Programme (Instrument for Nuclear Safety Co-operation). Within the first phase of this project, DBE Technology GmbH prepared a safety analysis report of the facility in its current state (SAR) and a preliminary safety analysis report (PSAR) for a future extended facility based on the planned enlargement. In addition to a detailed mathematical model, also simplified models have been developed to verify results of the former one and enhance confidence in the results. Comparison of the results show that - depending on the boundary conditions - simplifications like modeling the multi trench repository as one generic trench might have very limited influence on the overall results compared to the general uncertainties associated with respective long-term calculations. In addition to their value in regard to verification of more complex models which is important to increase confidence in the overall results, such simplified models can also offer the possibility to carry out time consuming calculations like probabilistic calculations or detailed sensitivity analysis in an economic manner. (authors)

Planning and Design Considerations for Geological Repository Programmes of Radioactive Waste

International Nuclear Information System (INIS)

2014-11-01

Disposal in a geological repository is the generally accepted solution for the long term management of high level and/or long lived radioactive wastes, in line with the general principles defined in the IAEA Safety Fundamentals. This publication presents practical information on the way a geological repository programme for radioactive waste could be defined and planned, with special attention to all aspects having an impact on the timing. Country specific examples for repository development phases are provided, based on actual experiences from Member States

PA/SA for Slovenian LILW repository

International Nuclear Information System (INIS)

Zeleznik, N.; Mele, I.

1999-01-01

The RAO Agency started with a new site selection procedure in 1996. As part of the preparational work for the new disposal facility, tools for assessment of the specific disposal concept influence on the environment and human has to be developed. Therefore the Slovenian assessment team that has been organized, joined the IAEAs ISAM programme, in which different approaches to performance and safety assessment were applied to safety cases. As part of the ISAM individual (national) safety cases. The RAO Agency, together with other Slovenian Inst.ions, performed the preliminary performance assessment of the Slovenian LILW repository for generic site location. The method and the results of the safety case are presented in this paper.(author)

The role of waste package specifications as a forerunner to ILW repository conditions for acceptance

International Nuclear Information System (INIS)

Barlow, S.V.; Palmer, J.D.

1998-01-01

In the absence of a finalized repository site, design or associated safety case, Nirex is not in a position to issue conditions for acceptance. Nirex has therefore developed a strategy which facilitates packaging of intermediate level waste by providing guidance through waste package specifications, supported by the formal assessment of specific packaging proposals on a case-by-case basis. The waste package specifications are comprehensive and cover all aspects of the waste package including dimensions and other key features, performance standards, wasteform, quality assurance, and data recording requirements. The waste package specifications will be subject to periodic review as repository design and safety cases are finalized and will progressively become site- and design-specific. The waste package specifications will eventually form the basis for conditions for acceptance. The strategy described in this paper has been successfully followed by Nirex and customers for the past ten years and has permitted wastes to be packaged for a deep repository with confidence in the absence of a finalized site and safety cases for the repository. Because the process has its basis in a generic repository concept, it remains robust, despite the increased uncertainty following the March 1997 Secretary of State's decision, as to the siting and time-scale of a deep waste repository, and continues to be an important component of the UK's waste management strategy. (author)

Review of Quality Assurance in SKB's Repository Research Experiments

Energy Technology Data Exchange (ETDEWEB)

Hicks, T.W. [Galson Sciences Ltd, 5 Grosvenor House, Melton Road, Oakham(United Kingdom)

2007-01-15

SKB is preparing licence applications for a spent nuclear fuel encapsulation plant and repository which will be supported by the SR-Site safety report. A separate safety report, SR-Can, has been produced by SKB in preparation for the SR-Site report. SKI is in the process of reviewing the SR-Can safety report. In preparation for this review, and with a view to building confidence in SKB's research activities and understanding SKB's handling of data and other information, SKI has examined SKB's application of QA measures in the management and conduct of repository research and development projects that support the SR-Can safety assessment. These preliminary investigations will serve to support the preparation of more detailed quality and technical audits of SKB's repository safety assessment after the submission of a licence application. SKI's approach to this QA review is based on the consideration of quality-affecting aspects of a selection of SKB's research and development activities. As part of this review, SKI identified the need to examine quality-related aspects of some of the many experiments and investigations that form part of SKB's repository research programme. This report presents the findings of such a review, focusing on experiments concerned with the properties and performance of the engineered barrier system. First, in order to establish a broad understanding of QA requirements for repository scientific investigations, QA procedures implemented in the management of research and development activities for the low-level radioactive waste repository near Drigg in the UK and the Waste Isolation Pilot Plant and Yucca Mountain repository projects in the US were studied. The QA procedures for experiments and tests undertaken in these projects were compared with those implemented by SKB. Key findings are: QA programmes have been implemented for each repository development programme in response to regulatory requirements

Standardized safety management of AP1000 nuclear power plant

International Nuclear Information System (INIS)

Li Xingwen; Cao Zhiqiang; Cong Jiuyuan

2011-01-01

In 2002, China published and implemented the Law of the People's Republic of China on Work Safety and promulgated a series of guidelines and policies, which strengthened the safety management supervision. Standardization of safety, as another important step on safety supervision, comes after safety assesment and safety production licensing system, is also a permanent solution. Standardization of safety is a strategic, long term and fundamental work, which is also the basic access to achieving scientific safety management and increasing the inherent safety of an enterprise. Haiyang AP1000 nuclear power plant, adopting the modularized, 'open-top' and parallel construction means, overturned the traditional construction theory of installation work comes after the civil work and greatly shorten the construction period. At the same time, the notable increase of oversize module transportation and lifting and parallel construction raises higher demands for safety management. This article combines the characteristics and difficulties of safety management for Haiyang AP1000 nuclear power plant, puts forward ideas and methods for standardized safety management, and could also serve as reference to the safety management for other AP1000 projects. (authors)

Aspects on the gas generation and migration in repositories for high level waste in salt formations

International Nuclear Information System (INIS)

Ruebel, Andre; Buhmann, Dieter; Meleshyn, Artur; Moenig, Joerg; Spiessl, Sabine

2013-07-01

In a deep geological repository for high-level waste, gases may be produced during the post-closure phase by several processes. The generated gases can potentially affect safety relevant features and processes of the repository, like the barrier integrity, the transport of liquids and gases in the repository and the release of gaseous radionuclides from the repository into the biosphere. German long-term safety assessments for repositories for high-level waste in salt which were performed prior 2010 did not explicitly consider gas transport and the consequences from release of volatile radionuclides. Selected aspects of the generation and migration of gases in repositories for high-level waste in a salt formation are studied in this report from the viewpoint of the performance assessment. The knowledge on the availability of water in the repository, in particular the migration of salt rock internal fluids in the temperature field of the radioactive waste repository towards the emplacement drifts, was compiled and the amount of water was roughly estimated. Two other processes studied in this report are on the one hand the release of gaseous radionuclides from the repository and their potential impact in the biosphere and on the other hand the transport of gases along the drifts and shafts of the repository and their interaction with the fluid flow. The results presented show that there is some gas production expected to occur in the repository due to corrosion of container material from water disposed of with the backfill and inflowing from the host rock during the thermal phase. If not dedicated gas storage areas are foreseen in the repository concept, these gases might exceed the storage capacity for gases in the repository. Consequently, an outflow of gases from the repository could occur. If there are failed containers for spent fuel, radioactive gases might be released from the containers into the gas space of the backfill and subsequently transported together

Deep repository for long-lived low- and intermediate-level waste in Sweden (SFL 3-5): An international peer review of SKB 's preliminary safety assessment

Energy Technology Data Exchange (ETDEWEB)

Chapman, N. [QuantiSci Ltd, Melton Mowbray (United Kingdom); Apted, M. [Monitor Scientific, Denver, CO (United States); Glasser, F. [Univ. of Aberdeen (United Kingdom). Dept. of Chemistry; Kessler, J. [EPRI, Inc., Palo Alto, CA (United States); Voss, C. [US Geological Survey, Reston, VA (United States)

2000-10-01

The SKB safety assessment of the SFL 3-5 repository (the planned deep repository for long-lived low- and intermediate level waste) can be read in two contexts: as a preliminary evaluation of the performance and design options for a repository that will not be required for perhaps forty years; or as an evaluation of a repository that might need to be sited together with the SFL 2 spent fuel repository, and whose nature and performance might thus need to be understood to a level that can be used to make wider programmatic decisions during the next five years. These two 'assessment contexts' are quite different, and an overarching issue is the fact that it was not clear to the review team which view to take. Apparently, SKB would tend towards the first context. However, it is not at all apparent to the reviewers why the second context should not be the predominant driver in the near future. The review team notes that the SFL 3-5 repository, as modelled by SKB, gives rise to potentially perceptible radionuclide releases to the environment on a timescale of hundreds of years after closure. This is in contrast to the SR 97 assessment for the SFL 2 spent fuel repository, which base scenario predicts no releases over a million year timescale. It is clear that according to SKB's SR97 and SFL3-5 analyses, for co-located facilities, it is this repository that has the potential for real radiological impacts in the immediate future. An initial recommendation from the review, is that SKB and the regulatory authorities consider which context is appropriate to the current status of the Swedish programme. This is important, because an overall impression of the reviewers is that the analysis would not be 'fit for purpose' if it were needed to assist with decision-making by SKB or the regulatory agencies. There are too many unanswered questions, and the overall impression of the safety concept is one of some fragility. Because there is no real design basis

Groundwater flow modelling of an abandoned partially open repository

Energy Technology Data Exchange (ETDEWEB)

Bockgaard, Niclas (Golder Associates AB (Sweden))

2010-12-15

As a part of the license application, according to the nuclear activities act, for a final repository for spent nuclear fuel at Forsmark, the Swedish Nuclear Fuel and Waste Management Company (SKB) has undertaken a series of groundwater flow modelling studies. These represent time periods with different hydraulic conditions and the simulations carried out contribute to the overall evaluation of the repository design and long-term radiological safety. The modelling study presented here serves as an input for analyses of so-called future human actions that may affect the repository. The objective of the work was to investigate the hydraulic influence of an abandoned partially open repository. The intention was to illustrate a pessimistic scenario of the effect of open tunnels in comparison to the reference closure of the repository. The effects of open tunnels were studied for two situations with different boundary conditions: A 'temperate' case with present-day boundary conditions and a generic future 'glacial' case with an ice sheet covering the repository. The results were summarized in the form of analyses of flow in and out from open tunnels, the effect on hydraulic head and flow in the surrounding rock volume, and transport performance measures of flow paths from the repository to surface

Groundwater flow modelling of an abandoned partially open repository

International Nuclear Information System (INIS)

Bockgaard, Niclas

2010-12-01

As a part of the license application, according to the nuclear activities act, for a final repository for spent nuclear fuel at Forsmark, the Swedish Nuclear Fuel and Waste Management Company (SKB) has undertaken a series of groundwater flow modelling studies. These represent time periods with different hydraulic conditions and the simulations carried out contribute to the overall evaluation of the repository design and long-term radiological safety. The modelling study presented here serves as an input for analyses of so-called future human actions that may affect the repository. The objective of the work was to investigate the hydraulic influence of an abandoned partially open repository. The intention was to illustrate a pessimistic scenario of the effect of open tunnels in comparison to the reference closure of the repository. The effects of open tunnels were studied for two situations with different boundary conditions: A 'temperate' case with present-day boundary conditions and a generic future 'glacial' case with an ice sheet covering the repository. The results were summarized in the form of analyses of flow in and out from open tunnels, the effect on hydraulic head and flow in the surrounding rock volume, and transport performance measures of flow paths from the repository to surface

Safety Problems of Disposal of Disused Sealed Sources in the Baldone Near Surface Repository

International Nuclear Information System (INIS)

Dreimanis, A.

2003-01-01

Current Latvian regulations encourage re-export of DSS, however, up to now in the repository has been disposed a lot of DSS. Baldone repository has received RW (mainly DSS) from ∼ 300 Latvia objects and from Kaliningrad region (from 1964 to 1973). DSS cause the major part of activity of the repository - from the total activity in facility ∼ 400 TBq more than 300 TBq is due to DSS. A long term Safety Assessment (SA) of the Baldone Radon-type near surface disposal facility has been performed by the consortium CASSIOPEE and recommendations are given. The waste disposal system consists of: vaults 1-6 (closed, as permanent disposal site), vault 7 (as a long-term retrievable storage site); 2 main components of the engineered barriers: the capping system and the vaults. Vaults 1 and 3-6 (of re-fabricated concrete elements, closed with reinforced concrete slabs, covered with hydro isolating layer completed by a sand/ soil layer). The vault No.7 (ten 130 m 3 underground concrete walled storage cells, protected from the weather by the building. These 10 underground storage cells are adjacent (2x5) with additional concrete walls in some of them. The cells are covered by 40 cm concrete slabs placed side to side. The assessment context and development of scenarios are presented in the paper. The results from the SA are presented. For the waste pathway scenarios - the resulting doses - ∼ 4 mSv/y at 30 y in the current status without cover, justify the implantation of a cover for the closure period of the repository. For air pathway scenarios - the basic dose target 1 mSv/y (for public) is not satisfied for all scenarios; the resulting effective dose for the on-site 'Western' residential scenario for the vaults with highest content of DSS: vault 3 - 303 mSv/y; vault 7 - 204 mSv/y. The main recommendations from the SA are: General Advice - to dispose sources with half-lives < 5,3 y; for the spent sources - to build a new long-term storage; 2. To move DSS from Vault 7 to the

Radioactive waste transport to a Nirex deep repository

International Nuclear Information System (INIS)

Bennett, D.; Appleton, P.R.; Eastman, C.R.

1989-01-01

Nirex is addressing the transport of radioactive wastes, repository construction materials, personnel and spoil as part of their development of a deep repository. An integrated transport system will be developed for wastes which may involve, road, rail and sea transport. The possible application and the scale of operation of the transport system is described. Environmental impact assessments will be carried out, and the proposed approach to these is described. A methodology for the assessment of transport safety has been established and the results of a preliminary assessment are given. (author)

Modelling of far-field gas migration from a deep radioactive waste repository

International Nuclear Information System (INIS)

Rodwell, W.R.; Nash, P.J.

1992-01-01

In assessing the post-closure safety of a deep radioactive waste repository, it is necessary to show that gas generated within the repository can migrate away, through the far-field geology, without affecting repository safety. This paper discusses the contribution of various mechanisms to gas migration through the far field; for example, diffusion of dissolved gas versus gas-phase movement, and bubble flow versus formation of a connected gas stream. It outlines different approaches to modelling gas movement from a repository, with simple semi-analytical models furnishing physical insights into the factors controlling gas migration in the absence of directly applicable experimental data, and more comprehensive numerical computations allowing the exploration of more detailed behaviour when appropriate data is obtained. If gas can induce groundwater movement, this could accelerate the transport of water-borne contaminants. Processes by which this could occur are noted, and the current status of work on possible effects of gas migration on groundwater movement in fractured hard rocks is indicated. 14 refs., 4 figs

The general situation of clay site for high-level waste geological disposal repository

International Nuclear Information System (INIS)

Wang Changxuan; Liu Xiaodong; Liu Pinghui

2008-01-01

Host medium is vitally important for safety of high-level radiaoactive waste (HLW) geological disposal. Clay, as host media of geological repository of HLW, has received greater attention for its inherent advantages. This paper summarizes IAEA and OECD/NEA's some safety guides on site selection and briefly introduces the process of the site selection, their studies and the characteristics of the clay formations in Switz-erland, France and Belgian. Based on these analyses, some suggestions are made to China's HLW repository clay site selection. (authors)

Development of the JNC geological disposal technical information integration system subjected for repository design and safety assessment

International Nuclear Information System (INIS)

Ishihara, Yoshinao; Ito, Takashi; Kobayashi, Shigeki; Neyama, Atsushi

2004-02-01

On this work, system manufacture about disposal technology and safety assessment field was performed towards construction of the JNC Geological Disposal Technical Information Integration System which systematized three fields of technical information acquired in investigation (site characteristic investigation) of geology environmental conditions, disposal technology (design of deep repository), and performance/safety assessment. The technical information database managed focusing on the technical information concerning individual research of an examination, analysis, etc. and the parameter set database managed focusing on the set up data set used in case of comprehensive evaluation are examined. In order to support and promote share and use of the technical information registered and managed by the database, utility functions, such as a technical information registration function, technical information search/browse function, analysis support function, and visualization function, are considered, and the system realized in these functions is built. The built system is installed in the server of JNC, and the functional check examination is carried out. (author)

Archaeological and natural analogs for the safety assessment of radioactive waste repositories

International Nuclear Information System (INIS)

Evstatiev, D.; Gergova, D.; Vachev, B.

2004-01-01

The safety assessment of surface repositories for low and intermediate level radioactive wastes (LILW) is based on scenarios assuming that water infiltration through the protective embankment increases with time, starting from 10% in the beginning and reaching 100% of the precipitation sum after 200-300 years. It is considered that this embankment will be destroyed by the atmospheric factors during the next centuries and that further on the concrete containers and repository chambers might fall apart, resulting finally in collapse of the protective barriers and exposing LILW to the direct impact of precipitation. These assumptions are rather conservative and for this reason the geological and anthropogenic analogs are of special interest as examples providing information about the real role of protective barriers of LILW repositories as well as about real migration processes occurring in the geoenvironment. Bulgaria could make a contribution to these investigations, since its territory is rich in ancient and medieval structures, tumuli from Thracian times and Old Bulgarian fortification banks, Roman and Medieval fortresses that might be regarded as analogs of the protective barriers of LILW repositories. A project will be launched in 2005 with objectives focused on this problem. The present report makes an analysis of the available information and states the tasks of the project. The ancient worldwide embankment practices in Egypt, China, Great Britain, Scythia, Thrace, Macedonia and other countries is briefly described. The Thracian tumuli in Bulgaria that reach the number of about 60 000 and the age of 2500 years. Their maximum height is 30 m, but the average height is between 7 and 10m. Their slope inclination is between 22 0 and 30 0 . Stone tombs had been found in some of the tumuli with excellently preserved frescoes, bas-reliefs and other articles of Thracian material culture. The horizontal thin layers of crushed limestone situated on top of the tombs

Opportunities for national repositories to resolve security challenges of past, present and future nuclear eras

International Nuclear Information System (INIS)

Sanders, T.L.

2005-01-01

With the global nuclear picture becoming even more complex, the role of repositories in accomplishing arms control, homeland security, and proliferation prevention goals has moved to front and center. Evolving repository infrastructures offer outstanding opportunities for illustrating advanced approaches for managing these risks. The traditional defense-in-depth concepts used to manage fuel cycle safety and protect nuclear materials in the U.S. and other countries could also be established as a framework for developing hardened, secure, and proliferation resistant material infrastructures including disposal systems. This analysis concept has been effective in establishing the safety basis for nuclear fuel cycles, reactors, and nuclear waste repositories. The concept results in the balanced use of multiple, diverse barriers to prevent the occurrence of undesired events such as radioactive releases from a safety perspective, or materials theft from a physical protection perspective. (author)

Understanding large scale groundwater flow to aid in repository siting

International Nuclear Information System (INIS)

Davison, C.C.; Brown, A.; Gascoyne, M.; Stevenson, D.R.; Ophori, D.U.

1996-01-01

Atomic Energy of Canada Limited (AECL) with support from Ontario Hydro has developed a concept for the safe disposal of Canada's nuclear fuel waste in a deep (500 to 1000 m) mined repository in plutonic rocks of the Canadian Shield. The disposal concept involves the use of multiple engineered and natural barriers to ensure long-term safety. The geosphere, comprised of the enclosing rock mass and the groundwater which occurs in cracks and pores in the rock, is expected to serve as an important natural barrier to the release and migration of wastes from the engineered repository. Although knowledge of the physical and chemical characteristics of the groundwater in the rock at potential repository sites is needed to help design the engineered barriers of the repository it can also be used to aid in repository siting, to take greater advantage of natural conditions in the geosphere to enhance its role as a barrier in the overall disposal system

Efficacy of backfilling and other engineered barriers in a radioactive waste repository in salt

International Nuclear Information System (INIS)

Claiborne, H.C.

1982-09-01

In the United States, investigation of potential host geologic formations was expanded in 1975 to include hard rocks. Potential groundwater intrusion is leading to very conservative and expensive waste package designs. Recent studies have concluded that incentives for engineered barriers and 1000-year canisters probably do not exist for reasonable breach scenarios. The assumption that multibarriers will significantly increase the safety margin is also questioned. Use of a bentonite backfill for surrounding a canister of exotic materials was developed in Sweden and is being considered in the US. The expectation that bentonite will remain essentially unchanged for hundreds of years for US repository designs may be unrealistic. In addition, thick bentonite backfills will increase the canister surface temperature and add much more water around the canister. The use of desiccant materials, such as CaO or MgO, for backfilling seems to be a better method of protecting the canister. An argument can also be made for not using backfill material in salt repositories since the 30-cm-thick space will provide for hole closure for many years and will promote heat transfer via natural convection. It is concluded that expensive safety systems are being considered for repository designs that do not necessarily increase the safety margin. It is recommended that the safety systems for waste repositories in different geologic media be addressed individually and that cost-benefit analyses be performed

The duration of the institutional controls on the low and intermediate level waste repositories

International Nuclear Information System (INIS)

Yang Jie; Li Yang; Liu Yafang; Lian Bing; Zhao Yangjun; Chen Hailong; Gu Zhijie

2014-01-01

Appropriate institutional controls are put in place prior to repository closure. Such controls can guarantee the long term safety of the repository. Today there is no clear standard on how to determine the institutional control period. This paper tries to give possible factors and activities of the institutional controls on the low and intermediate level waste repositories, and makes some suggestions on the institutional controls in our country. (authors)

Safety case approach for a KBS-3 type repository in crystalline rock

International Nuclear Information System (INIS)

Pastina, Barbara; Lehikoinen, Jarmo; Puigdomenech, Ignasi

2012-01-01

Barbara Pastina of Saanio and Riekkola described the approach to considering cementitious materials in a safety case for a KBS-3 repository in Finland. In this concept, cements will be used predominantly as tunnel plugs and seals. Part of the Finnish approach has involved identifying the cement-related FEPs. For example, FEPs representing the effects of cement on spent fuel, on the canister and on radionuclide transport include: - Fuel matrix dissolution at high pH. - Copper corrosion at high pH. - Radionuclide speciation and solubility at high pH. - Radionuclide sorption and diffusion at high pH. - Radionuclide transport due to organic materials (e.g. super-plasticisers). - Colloid formation at a high pH plume front. FEPs representing the effects of cement on bentonite in the buffer and backfill include: - Potential changes in swelling pressure due to mass loss, decrease in clay density, and precipitation of secondary minerals. - Potential cracking and increase of hydraulic conductivity due to cementation. - Increase of the cation exchange capacity due to the loss of silica from the montmorillonite structure. Amongst the cement-related FEPs, the main concerns are related to effects on the performance of the bentonite buffer. Cement-bentonite interactions are complex, there are few experimental data, and there are significant modelling uncertainties (e.g. limited knowledge about the reactions that may occur and their rates, and the effects of temperature). Accepting the existence of various uncertainties, preliminary modelling studies performed using the TOUGHREACT code illustrate the potential for porosity reduction and clogging of porosity in bentonite affected by cementitious pore waters. The modelling also suggests that that the high pH of the pore waters moving from the cementitious materials into the bentonite may be rapidly lowered as a result of reactions with the bentonite close to the cement-bentonite interface. Taking account of the various research and

Microbial processes in a clay repository

Energy Technology Data Exchange (ETDEWEB)

Canniere, Pierre de [Federal Agency of Nuclear Control (FANC), Brussels (Belgium); Meleshyn, Artur [Gesellschaft fuer Anlagen- und Reaktorsicherheit mbH (GRS), Braunschweig (Germany)

2013-07-01

The safety of a deep geologic repository (DGR) for nuclear waste must be ensured for geological times exceeding human imagination taking into account large uncertainties. The long-term effects of complex biogeochemical processes potentially affecting the integrity and the long-term safety of engineered barriers might still be unknown. The aim of this presentation is to give a general overview of some microbial processes which have contributed to shape the Earth since probably billions of years and whose unexpected consequences for nuclear waste disposal should be appropriately tackled. (orig.)

Safety and performance indicators for the assessment of long-term safety of deep geological disposal of radioactive waste

International Nuclear Information System (INIS)

Hugi, M.; Schneider, J.W.; Dorp, F. van; Zuidema, P.

2005-01-01

The evaluation of the ability to isolate radioactive waste and the assessment of the long-term safety of a deep geological repository is usually done in terms of the calculated dose and/or risk for an average individual of the population which is potentially most affected by the potential impacts of the repository. At present, various countries and international organisations are developing so-called complementary indicators to supplement such calculations. These indicators are called ''safety indicators'' if they refer to the safety of the whole repository system; if they address the isolation capability of individual system components or the whole system from a more technical perspective, they are called ''performance indicators''. The need for complementary indicators follows from the long time frames which characterise the safety assessment of a geological repository, and the corresponding uncertainty of the calculated radiation dose. The main reason for these uncertainties is associated with the uncertain long-term prognosis of the surface environment and the related human behaviour. (orig.)

Safeguards for geological repositories

International Nuclear Information System (INIS)

Fattah, A.

2000-01-01

substantially large. Change in social, economic, environmental and other scenarios might demand recovery of nuclear and other material from the repository sometime in the future. To this end, the Department of Safeguards has developed a policy paper to guide the planner, designer and operator to incorporate safeguards related features, as appropriate. In parallel, a programme for the Development of Safeguards for Final Disposal of Spent Fuel in Geological Repositories (SAGOR) was launched to foster technological advancement. The mission of SAGOR has been to ensure that the safeguards systems developed for the final disposal of spent fuel effectively meet the objectives of IAEA safeguards, optimise IAEA resources, and make best use of existing technologies while still meeting the requirements for safety and environmental protection. (author)

Deep repository for long-lived low- and intermediate-level waste in Sweden (SFL 3-5): An international peer review of SKB 's preliminary safety assessment

International Nuclear Information System (INIS)

Chapman, N.; Apted, M.; Glasser, F.; Voss, C.

2000-10-01

The SKB safety assessment of the SFL 3-5 repository (the planned deep repository for long-lived low- and intermediate level waste) can be read in two contexts: as a preliminary evaluation of the performance and design options for a repository that will not be required for perhaps forty years; or as an evaluation of a repository that might need to be sited together with the SFL 2 spent fuel repository, and whose nature and performance might thus need to be understood to a level that can be used to make wider programmatic decisions during the next five years. These two 'assessment contexts' are quite different, and an overarching issue is the fact that it was not clear to the review team which view to take. Apparently, SKB would tend towards the first context. However, it is not at all apparent to the reviewers why the second context should not be the predominant driver in the near future. The review team notes that the SFL 3-5 repository, as modelled by SKB, gives rise to potentially perceptible radionuclide releases to the environment on a timescale of hundreds of years after closure. This is in contrast to the SR 97 assessment for the SFL 2 spent fuel repository, which base scenario predicts no releases over a million year timescale. It is clear that according to SKB's SR97 and SFL3-5 analyses, for co-located facilities, it is this repository that has the potential for real radiological impacts in the immediate future. An initial recommendation from the review, is that SKB and the regulatory authorities consider which context is appropriate to the current status of the Swedish programme. This is important, because an overall impression of the reviewers is that the analysis would not be 'fit for purpose' if it were needed to assist with decision-making by SKB or the regulatory agencies. There are too many unanswered questions, and the overall impression of the safety concept is one of some fragility. Because there is no real design basis presented, no thorough

Influence of the design temperature on long-term safety of a salt dome repository

International Nuclear Information System (INIS)

Buhmann, D.; Brenner, J.; Storck, R.

1993-03-01

All studies made so far within the framwork of the mixed concept system analysis proceeded from a design temperature of the mine structure of 200 C. The concept based on a design temperature of 150 C was aimed at studying whether it made sense to maintain lower temperatures, if necessary. Deterministic and probabilistic calculations were made in order to determine the influence of the lower design temperature on long-term safety. The calculations were based on concept A of Joint Borehole and Gallery Storage. Assuming reference values of the input parameters, the deterministic calculations do not produce any radionuclide release from the mine structure. If, however, one assumes a lower rate for rock convergence, radionuclides are released at maximum dose rates of about 3.10 -5 Sv/a. Even a larger volume of limited brine inclusions may lead to radionuclide releases, in that case with dose commitments of the order of magnitude of 1.10 -5 Sv/a. The probabilistic calculations show that a design temperature of 150 C for long-term safety is less favourable than a higher design temperature. The share of simulations in the probabilistic calculations with a radionuclide release, and the expected value of dose commitment, are almost double as high as in the concept based on 200 C design temperature. Thus a higher design temperature is preferable with regard to the long-term safety of a salt repository. The most important parameters concerning dose commitment are the volume of limited brine inclusions, the convergence rate, and the permeability of barriers and backfilling rock. (orig./HP) [de

The Microbiology of Subsurface, Salt-Based Nuclear Waste Repositories: Using Microbial Ecology, Bioenergetics, and Projected Conditions to Help Predict Microbial Effects on Repository Performance

International Nuclear Information System (INIS)

Swanson, Juliet S.; Cherkouk, Andrea; Arnold, Thuro; Meleshyn, Artur; Reed, Donald T.

2016-01-01

This report summarizes the potential role of microorganisms in salt-based nuclear waste repositories using available information on the microbial ecology of hypersaline environments, the bioenergetics of survival under high ionic strength conditions, and ''repository microbiology'' related studies. In areas where microbial activity is in question, there may be a need to shift the research focus toward feasibility studies rather than studies that generate actual input for performance assessments. In areas where activity is not necessary to affect performance (e.g., biocolloid transport), repository-relevant data should be generated. Both approaches will lend a realistic perspective to a safety case/performance scenario that will most likely underscore the conservative value of that case.

The Microbiology of Subsurface, Salt-Based Nuclear Waste Repositories: Using Microbial Ecology, Bioenergetics, and Projected Conditions to Help Predict Microbial Effects on Repository Performance

Energy Technology Data Exchange (ETDEWEB)

Swanson, Juliet S. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Cherkouk, Andrea [Helmholtz-Zentrum Dresden-Rossendorf, Rossendorf (Germany); Arnold, Thuro [Helmholtz-Zentrum Dresden-Rossendorf, Rossendorf (Germany); Meleshyn, Artur [Gesellschaft fur Anlagen und Reaktorsicherheit, Braunschweig (Germany); Reed, Donald T. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2016-11-17

This report summarizes the potential role of microorganisms in salt-based nuclear waste repositories using available information on the microbial ecology of hypersaline environments, the bioenergetics of survival under high ionic strength conditions, and “repository microbiology” related studies. In areas where microbial activity is in question, there may be a need to shift the research focus toward feasibility studies rather than studies that generate actual input for performance assessments. In areas where activity is not necessary to affect performance (e.g., biocolloid transport), repository-relevant data should be generated. Both approaches will lend a realistic perspective to a safety case/performance scenario that will most likely underscore the conservative value of that case.

Abadia de Goias repository: design conception; Repositorio de Abadia de Goias: concepcao de projeto

Energy Technology Data Exchange (ETDEWEB)

Martin Alves, Antonio Sergio de; Santos, Cicero Durval Pacifici dos; Passos, Erivaldo Mario dos; Coutinho, Fernando Paulo Millen [NUCLEN, Rio de Janeiro, RJ (Brazil)

1995-12-31

In this paper have been presented the criteria, the methodologies and the parameters that were utilized for the design of Abadia de Goias Repository. Hereby the purpose is to show in a succinct way the know how that has been acquired for the design of a LLW and ILW repository. This paper presents information and details concerning to the various phases of the design, beginning with the data collecting activity, the safety analysis elaboration up to the definition of the final concept of the repository and of the required infrastructure work. The safety analysis, based on the Cs-137 migration through the groundwater, made possible at first to determine the places of the repository site where the population is not allowed to drill wells. The analysis allowed also the institutional control period calculation based on the maximum concentration of Cs-137 in the aquifer as well as in the intrusion models. (author). 16 refs., 3 figs., 1 tab.

Quality assurance for safety in the radioactive waste management: a quality assurance system in Novi Han radioactive waste repository

International Nuclear Information System (INIS)

Petrova, A.; Kolev, I.

2000-01-01

Novi Han Radioactive Waste Repository (RWR) is still the only place in Bulgaria for storage of low and intermediate level radioactive waste. It is necessary to establish and maintain a Quality Assurance (QA) system to ensure that the RWR can be operated safely with regard to the health and safety of the general public and site personnel. A QA system has to establish the basic requirements for quality assurance in order to enhance nuclear safety by continuously improving the methods employed to achieve quality. It is envisaged that the QA system for the Novi Han RWR will cover the operation and maintenance of the radioactive waste disposal facilities, the radiation protection and monitoring of the site, as well as the scientific and technology development aspects. The functions of the Novi Han RWR presume the availability of an environmental management system. It is appropriate to establish a QA system based on the requirements of the ISO Standards 9001 and 14000, using the recommendations of the IAEA (Quality assurance for safety in NPPs and other nuclear installations, code and safety guides Q1-Q14). (authors)

SUBSURFACE REPOSITORY INTEGRATED CONTROL SYSTEM DESIGN

International Nuclear Information System (INIS)

Randle, D.C.

2000-01-01

, controlled, and interfaced (Section 6.2). (3) Develop a preliminary design for the overall Subsurface Repository Integrated Control System functional architecture and graphically depict the operational features of this design through a series of control system functional block diagrams (Section 6.2). (4) Develop a physical architecture that presents a viable yet preliminary physical implementation for the Subsurface Repository Integrated Control System functional architecture (Section 6.3). (5) Develop an initial concept for an overall subsurface data communications network that can be used to integrate the various control systems comprising the Subsurface Repository Integrated Control System (Section 6.4). (6) Develop a preliminary central control room design for the Subsurface Repository Integrated Control System (Section 6.5). (7) Identify and discuss the general safety-related issues and design strategies with respect to development of the Subsurface Repository Integrated Control System (Section 6.6). (8) Discuss plans for the Subsurface Repository Integrated Control System's response to off-normal operations (Section 6.7). (9) Discuss plans and strategies for developing software for the Subsurface Repository Integrated Control System (Section 6.8)

Preliminary design of the repository. Stage 2

International Nuclear Information System (INIS)

Saanio, T.; Kirkkomaeki, T.; Keto, P.; Kukkola, T.; Raiko, H.

2007-04-01

Spent nuclear fuel from Finnish nuclear power plants will be disposed of in deep bedrock in Olkiluoto, Eurajoki. The repository is planned to be excavated at a depth of 400 - 500 metres. Access routes to the repository include a 1:10 inclined access tunnel, and vertical shafts. The fuel is encapsulated in the encapsulation plant above ground and transferred to the repository in the canister lift. Deposition tunnels, central tunnels and technical rooms are excavated at the disposal level. The canisters are deposited in deposition holes that are covered with bentonite blocks. The deposition holes are bored in the floors of the deposition tunnels. The central tunnel system consists of two parallel central tunnels that are inter-connected at certain distances. Two parallel central tunnels improve the fire safety of the rooms and also allow flexible backfilling and closing of the deposition tunnels in stages at the operational phase of the repository. An underground rock characterization facility, ONKALO, is excavated at the disposal level to support and confirm investigations carried out from above ground. ONKALO is designed so that it can later serve as part of the repository. ONKALO excavations were started in 2004. The repository will be excavated in the 2010s and operation will start in 2020. The fifth nuclear power unit makes the operational phase of the repository very long. Parts of the repository will be excavated and closed over the long operational period. The repository can be constructed at one or several levels. The one-storey alternative is the so-called reference alternative in this preliminary design report. The two-storey alternative is also taken into account in the ONKALO designs. The preliminary designs of the repository are presented as located in Olkiluoto. The location of the repository will be revised when more information on the bedrock has been gained. More detailed data of the circumstances will be obtained from above ground investigations

Preliminary design of the repository, stage 2

International Nuclear Information System (INIS)

Saanio, T.; Kirkkomaeki, T.; Keto, P.; Kukkola, T.; Raiko, H.

2007-01-01

Spent nuclear fuel from Finnish nuclear power plants will be disposed of in deep bedrock in Olkiluoto, Eurajoki. The repository is planned to be excavated at a depth of 400 - 500 metres. Access routes to the repository include a 1:10 inclined access tunnel, and vertical shafts. The fuel is encapsulated in the encapsulation plant above ground and transferred to the repository in the canister lift. Deposition tunnels, central tunnels and technical rooms are excavated at the disposal level. The canisters are deposited in deposition holes that are covered with bentonite blocks. The deposition holes are bored in the floors of the deposition tunnels. The central tunnel system consists of two parallel central tunnels that are inter-connected at certain distances. Two parallel central tunnels improve the fire safety of the rooms and also allow flexible backfilling and closing of the deposition tunnels in stages at the operational phase of the repository. An underground rock characterization facility, ONKALO, is excavated at the disposal level to support and confirm investigations carried out from above ground. ONKALO is designed so that it can later serve as part of the repository. ONKALO excavations were started in 2004. The repository will be excavated in the 2010s and operation will start in 2020. The fifth nuclear power unit makes the operational phase of the repository very long. Parts of the repository will be excavated and closed over the long operational period. The repository can be constructed at one or several levels. The one-storey alternative is the so-called reference alternative in this preliminary design report. The two-storey alternative is also taken into account in the ONKALO designs. The preliminary designs of the repository are presented as located in Olkiluoto. The location of the repository will be revised when more information on the bedrock has been gained. More detailed data of the circumstances will be obtained from above ground investigations

Gas generation and release from the VLJ repository

International Nuclear Information System (INIS)

Vieno, T.; Valkiainen, M.

1992-01-01

The VLJ repository is an underground disposal facility located at the Olkiluoto nuclear power plant site on the west coast of Finland. The repository will house low (LLW) and intermediate level radioactive wastes (MLW) from the TVO I and TVO II BWR's and the spent fuel interim store at Olkiluoto. The disposal rooms have been excavated at a depth of 60... 100 meters in the crystalline bedrock. They consist of two rock silos - one for the LLW and the other for MLW. Low level waste is usually packed in steel drums and steel boxes. Medium level wastes consists of bituminized resins in steel drums. Wastes packages are emplaced in concrete boxes before transportation into the repository. Low level wastes are emplaced in the shotcreted rock silo where no backfilling will used. For medium level wastes, a separate silo of reinforced concrete has been constructed inside the rock silo. No backfilling will be used inside the concrete silo and an opening will be made in the lid of the concrete silo for gas release. The microbial degradation of low level wastes is the principle gas generation process in the repository. The gas transport though the bedrock covering the repository is evaluated with the help of ground water flow study. It is recommended that the shotcrete lining on the ceiling of the repository cavern is partly removed before the final sealing of the repository. Provided that dissipation of gases from the disposal cavern into the rock can been assured, the overall effects of gas generation on the long-term safety of the repository are insignificant. 10 refs., 6 figs

Long term effects on potential repository sites: climatic and geomorphological changes

International Nuclear Information System (INIS)

Seddon, M.B.; Worsley, P.

1985-05-01

A study of the effects of climatic variability on the geomorphological processes operating on the landscape are important in the study of radioactive waste repository sites. The effects of glacial erosion and deposition are fundamental to an examination of repository safety, particularly in North Britain. Rates of climatic shift need to be examined. Predictive simulation models, based on a knowledge of past climatic events, for future global climates are proposed. (UK)

Criticality issues with highly enriched fuels in a repository environment

International Nuclear Information System (INIS)

Taylor, L.L.; Sanchez, L.C.; Rath, J.S.

1998-03-01

This paper presents preliminary analysis of a volcanic tuff repository containing a combination of low enrichment commercial spent nuclear fuels (SNF) and DOE-owned SNF packages. These SNFs were analyzed with respect to their criticality risks. Disposal of SNF packages containing significant fissile mass within a geologic repository must comply with current regulations relative to criticality safety during transportation and handling within operational facilities. However, once the repository is closed, the double contingency credits for criticality safety are subject to unremediable degradation, (e.g., water intrusion, continued presence of neutron absorbers in proximity to fissile material, and fissile material reconfiguration). The work presented in this paper focused on two attributes of criticality in a volcanic tuff repository for near-field and far-field scenarios: (1) scenario conditions necessary to have a criticality, and (2) consequences of a nuclear excursion that are components of risk. All criticality consequences are dependent upon eventual water intrusion into the repository and subsequent breach of the disposal package. Key criticality parameters necessary for a critical assembly are: (1) adequate thermal fissile mass, (2) adequate concentration of fissile material, (3) separation of neutron poison from fissile materials, and (4) sufficient neutron moderation (expressed in units of moderator to fissile atom ratios). Key results from this study indicated that the total energies released during a single excursion are minimal (comparable to those released in previous solution accidents), and the maximum frequency of occurrence is bounded by the saturation and temperature recycle times, thus resulting in small criticality risks

Natural analogues: studies of geological processes relevant to radioactive waste disposal in deep geological repositories

Energy Technology Data Exchange (ETDEWEB)

Russel, A.W. [Bedrock Geosciences, Auenstein (Switzerland); Reijonen, H.M. [Saanio and Rickkola Oy, Helsinki (Finland); McKinley, I.G. [MCM Consulting, Baden-Daettwil (Switzerland)

2015-06-15

The geological disposal of radioactive wastes is generally accepted to be the most practicable approach to handling the waste inventory built up from over 70 years accumulation of power production, research-medical-industrial and military wastes. Here, a brief overview of the approach to geological disposal is presented along with some information on repository design and the assessment of repository post-closure safety. One of the significant challenges for repository safety assessment is how to extrapolate the likely long-term (i.e. ten thousand to a million years) behaviour of the repository from the necessarily short term data from analytical laboratories and underground rock laboratories currently available. One approach, common to all fields of the geosciences, but also in such diverse fields as philosophy, biology, linguistics, law, etc., is to utilise the analogue argumentation methodology. For the specific case of radioactive waste management, the term 'natural analogue' has taken on a particular meaning associated with providing supporting arguments for a repository safety assessment. This approach is discussed here with a brief overview of how the study of natural (and, in particular, geological) systems can provide supporting information on the likely long-term evolution of a deep geological waste repository. The overall approach is discussed and some relevant examples are presented, including the use of uranium ore bodies to assess waste form stability, the investigation of native metals to define the longevity of waste containers and how natural clays can provide information on the stability of waste tunnel backfill material. (authors)

Natural analogues: studies of geological processes relevant to radioactive waste disposal in deep geological repositories

International Nuclear Information System (INIS)

Russel, A.W.; Reijonen, H.M.; McKinley, I.G.

2015-01-01

The geological disposal of radioactive wastes is generally accepted to be the most practicable approach to handling the waste inventory built up from over 70 years accumulation of power production, research-medical-industrial and military wastes. Here, a brief overview of the approach to geological disposal is presented along with some information on repository design and the assessment of repository post-closure safety. One of the significant challenges for repository safety assessment is how to extrapolate the likely long-term (i.e. ten thousand to a million years) behaviour of the repository from the necessarily short term data from analytical laboratories and underground rock laboratories currently available. One approach, common to all fields of the geosciences, but also in such diverse fields as philosophy, biology, linguistics, law, etc., is to utilise the analogue argumentation methodology. For the specific case of radioactive waste management, the term 'natural analogue' has taken on a particular meaning associated with providing supporting arguments for a repository safety assessment. This approach is discussed here with a brief overview of how the study of natural (and, in particular, geological) systems can provide supporting information on the likely long-term evolution of a deep geological waste repository. The overall approach is discussed and some relevant examples are presented, including the use of uranium ore bodies to assess waste form stability, the investigation of native metals to define the longevity of waste containers and how natural clays can provide information on the stability of waste tunnel backfill material. (authors)

Methodology of proving long-term safety of a salt dome repository with existing insecurities forming the background

International Nuclear Information System (INIS)

Storck, R.

1992-01-01

Existing methods to prove safety can consider the insecurities of input data within the framework of probabilistic analyses. The results of application calculations show that inspite of considerable band widths of input data the scattering widths of radiation exposures are comparably limited, and calculated radiation exposures are clearly below acceptable limits. Moreover it can be demonstrated that in the event of an assumed brine influx into the repository radionuclides are released only if parameter combinations are unfavourable. Therefore such incident in general does not have any radiological consequences. Insecurities in model approaches can be taken into consideration only partly so far by using alternative models, or indirectly through data insecurities. (orig./DG) [de

Review of computer models used for post closure safety assessment of nuclear waste repositories in the FRG

International Nuclear Information System (INIS)

Bogorinski, P.; Baltes, B.; Martens, K.H.

1987-01-01

In the FRG, disposal of nuclear wastes takes place in deep geologic formations. For longterm safety assessment of such a repository, groundwater transport provides a release scenario for the radionuclides to the biosphere. GRs reviewed a methodology that was implemented by the research group of PSE to simulate migration of radionuclides in the geosphere. The examination included the applicability of theoretical models, numerical experiments, comparison to results of diverse computer codes as well as experience from international intercomparison studies. The review concluded that the hydrological model may be applied to full extent unless density effects have to be considered whereas there are some restrictions in the use of the nuclide transport model

Judicial progress in Germany's nuclear waste disposal policy the Konrad repository decisions of 26 march 2007

International Nuclear Information System (INIS)

Kuhne, G.

2007-01-01

Its exists three final repository sites in Germany, an overview of the status of these sites is necessary to understand concretely the situation. Morsleben, is actually the only final repository for low and medium level radioactive waste. Gorleben is the site of a salt dome which is under exploration as an eventual repository for high level radioactive waste but the study is under a decree of exploration moratorium of at least three years, in order to allow for investigation into conceptual and safety issues. The Konrad ore mine is the site under preparation for a final repository for low and medium level radioactive waste. The present article will briefly address three aspects of the court reasoning: the legal character of the plan approval notice (act of discretion or strict execution of laws), the necessity of planning and safety aspects. (N.C.)

Radioactive waste repository of Cesium of Abadia de Goias. Construction and design; Repositorio de rejeitos radioativos de cesio - Abadia de Goias. Concepcao e projeto

Energy Technology Data Exchange (ETDEWEB)

Tranjan Filho, Alfredo [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil); Alves, Antonio Sergio de Martin; Santos, Cicero Durval Pacifici dos; Passos, Erivaldo Mario dos; Coutinho, Fernando Paulo Millen [NUCLEN Engenharia e Servicos S.A., Rio de Janeiro, RJ (Brazil)

1997-12-31

The main criteria, the methodology, the solutions and parameters that were utilized in the design of the Intermediate and Low Level Radioactive Waste Repository of Abadia de Goias are shortly described. The various design steps are analysed from the preparation of the Safety Analysis Report to the detailing engineering tasks. The safety analysis for the constructed repository had the goal of verifying the magnitude of radioecological impacts corresponding to idealized activity release scenarios, allowing also the possible effects of human intrusion in the repository. These safety studies are intrinsically connected to computer calculations envisaged to simulate the long term performance of the repository. (author) 18 refs., 7 figs., 7 tabs.

Radioactive Waste Repositories and Incentives to Local Communities

International Nuclear Information System (INIS)

Knapp, A.; Medakovic, S.

2008-01-01

Public acceptance of radioactive waste (RW) repository depends on various and often community-specific factors. Although radiological risk from a properly constructed low and intermediate level waste (LILW) repository is practically negligible, routine safety considerations will favor low populated areas and therefore probably underdeveloped communities. Repository acceptance in such communities is more likely to be facilitated by prospective benefits to local economy, such as infrastructure development and increased employment, as well as by dedicated financial incentives to the community. Direct financial compensation to the local community for acceptance of the repository has been considered in some documents in countries experienced in RW management, but it has not become a widely accepted practice. In Croatia, a possibility for such compensation is mentioned in the land use plan in conjunction with the prospective RW repository site. In Slovenia, the government has already specified the annual amount of 2.33 million euro as a compensation for 'limited land use' to be shared by local communities in the vicinity of the planned LILW repository during its operation. Applicability of the Slovenian compensations to the prospective joint Slovenian-Croatian repository is not yet clear, at least in the aspect of joint funding. The joint Slovenian-Croatian Decommissioning and LILW and SF management program for NPP Krsko from 2004 did conservatively include the compensations into the repository cost estimates, but that might not be retained in subsequent revisions of the Program. According to the agreement between governments of Slovenia and Croatia on the Nuclear power plant Krsko, Croatian side has no obligations to participate in 'public expenditures' introduced after the agreement, as would be the case of community compensations for LILW repository in Slovenia. Before further decisions on joint NPP Krsko waste management are made, including the issue of LILW

High level radioactive waste repositories. Task 3. Review of underground handling and emplacement. 1. Executive summary

Energy Technology Data Exchange (ETDEWEB)

1981-01-01

A review is presented of proposals for transport, handling and emplacement of high-level radioactive waste in an underground repository appropriate to the U.K. context, with particular reference to waste block size and configuration; self-shielded or partially-shielded block; stages of disposal; transport by road/rail to repository site; handling techniques within repository; emplacement in vertical holes or horizontal tunnels; repository access by adit, incline or shaft; conventional and radiological safety; costs; and major areas of uncertainty requiring research or development.

Siting regions for deep geological repositories. Nagra’s proposals for stage 3

International Nuclear Information System (INIS)

2014-01-01

This brochure published by the Swiss National Cooperative for the Disposal of Radioactive Waste (NAGRA) discusses the selection of sites for deep geological repositories for nuclear wastes in Switzerland. The procedure proposed for the selection process is explained. The four sites for possible repositories of high-level radioactive waste as well as for low and intermediate-level wastes are described and rated with respect to the various safety factors involved. The reasons for the long-term safety measures proposed and the geological barriers involved are discussed. The four proposals for depository sites are looked at in more detail. The paper is well illustrated with several diagrams and tables

Influence of climate on landscape characteristics in safety assessments of repositories for radioactive wastes.

Science.gov (United States)

Becker, J K; Lindborg, T; Thorne, M C

2014-12-01

In safety assessments of repositories for radioactive wastes, large spatial and temporal scales have to be considered when developing an approach to risk calculations. A wide range of different types of information may be required. Local to the site of interest, temperature and precipitation data may be used to determine the erosional regime (which may also be conditioned by the vegetation characteristics adopted, based both on climatic and other considerations). However, geomorphological changes may be governed by regional rather than local considerations, e.g. alteration of river base levels, river capture and drainage network reorganisation, or the progression of an ice sheet or valley glacier across the site. The regional climate is in turn governed by the global climate. In this work, a commentary is presented on the types of climate models that can be used to develop projections of climate change for use in post-closure radiological impact assessments of geological repositories for radioactive wastes. These models include both Atmosphere-Ocean General Circulation Models and Earth Models of Intermediate Complexity. The relevant outputs available from these models are identified and consideration is given to how these outputs may be used to inform projections of landscape development. Issues of spatial and temporal downscaling of climate model outputs to meet the requirements of local-scale landscape development modelling are also addressed. An example is given of how climate change and landscape development influence the radiological impact of radionuclides potentially released from the deep geological disposal facility for spent nuclear fuel that SKB (the Swedish Nuclear Fuel and Waste Management Company) proposes to construct at Forsmark, Sweden. Copyright © 2014 Elsevier Ltd. All rights reserved.

Predicted thermal and stress environments in the vicinity of repository openings

International Nuclear Information System (INIS)

Bauer, S.J.; Hardy, M.P.; Lin, M.

1991-01-01

An understanding of the thermal and stress environment in the vicinity of repository openings is important for preclosure performance considerations and worker health and safety considerations for the proposed high-level radioactive waste repository at Yucca Mountain. This paper presents the results of two and three dimensional numerical analyses which have determined the thermal and stress environments for typical repository openings. In general, it is predicted that openings close to heat sources attain high temperatures and experience a significant stress increase. Openings away from heat sources experience more uniform temperature changes and experience a stress change which results in part from a far-field thermal loading

Methodology applied in Cuba for siting, designing, and building a radioactive waste repository under safety conditions

International Nuclear Information System (INIS)

Orbera, L.; Peralta, J.L.; Franklin, R.; Gil, R.; Chales, G.; Rodriguez, A.

1993-01-01

The work presents the methodology used in Cuba for siting, designing, and building a radioactive waste repository safely. This methodology covers both the technical and socio-economic factors, as well as those of design and construction so as to have a safe siting for this kind of repository under Cuba especial condition. Applying this methodology will results in a safe repository

First Workshop on Design and Construction of Deep Repositories - Theme: Excavation through water-conducting major fracture zones

International Nuclear Information System (INIS)

Baeckblom, G.; Svemar, C.

1994-01-01

Final disposal of high-level nuclear waste has not yet been carried out in any country today. The concepts under development are all based on geological repositories, i.e., disposal at a sufficient depth below the surface to provide stable mechanical, hydrological and chemical conditions during the period the waste needs to be isolated from man. In the cases where crystalline bedrock is considered the proposed repository depths vary between 300-1000 m. The construction, operation and sealing of a deep geological repository must meet various criteria that in many respects are more detailed and more demanding than usual in underground construction projects today. The work shall be done so that occupational safety is ensured. The work also shall conform to whatever restrictions are necessary for ensuring pre-closure operational safety and post-closure long-term safety. March 1993 SKB arranged a two-day international workshop to discuss design and construction of repositories. Close to 40 participants from eight countries shared experiences regarding passage of major water-conducting fracture zones and other matters. This report summarizes the contributions to the workshop

Monitoring of IL/HL, LL waste repository in a clay formation: objectives, technical know-how, implementation strategy

International Nuclear Information System (INIS)

Grevoz, Arnaud; Mayer, Stefan; Dubois, Jean-Philippe

2005-01-01

In France, with regards to the monitoring of a reversible repository in a clay host formation, Andra utilised international references as input to develop its own strategy and program, e.g. IAEA TECDOC 1208 and DS-154 2004, NEA 2001 (reversibility and retrievability), EUR 21025 EN. Additionally, the French safety rule RFSIII.2.f, which aims to provide guidelines but is not a regulatory requirement, deals with 'general provisions concerning explorations'. The RFSIII.2.f distinguishes between exploration conducted from the surface, investigations to be carried out in the Underground Research Laboratory (URL) and monitoring of changes in the site while the repository is in operation. The monitoring programme responds to the three main motivations: 1. the respect of operational safety and regulatory requirements; 2. the acquisition of data to improve the understanding of processes and parameters underlying the long term safety assessment; and 3. reversibility. In that respect, Andra defines monitoring terminology by distinguishing two main activities: (i) observation (for scientific and engineering understanding and reversibility) and (ii) surveillance (related to operational and long term safety). Surveillance to contribute to the operational safety of a repository, e.g. surveillance of drift stability, fire and radiological hazards, is in line with other classical and nuclear operational safety principles and practices. Surveillance to contribute to long term safety can provide input to periodic re-evaluations of the safety analysis of a geologic repository, with a view of improving system understanding and confirming data in situ, prior to final closure of the repository. As such, it is in line with the usual regulatory guidance for Nuclear Power Plant (NPP) facilities. While it will not provide direct data on the long-term evolution of the system, it may provide some data, for example related to initial conditions or to an early transient phase, for

Preliminary waste acceptance requirements for the planned Konrad repository

International Nuclear Information System (INIS)

Warnecke, E.; Brennecke, P.

1987-01-01

The Physikalisch-Technische Bundesanstalt (PTB) has established Preliminary Waste Acceptance Requirements for the planned Konrad repository. These requirements were developed, in accordance with the Safety Criteria of the Reactor Safety Commission, with the help of a site specific safety assessment; they are under the reservation of the plan approval procedure, which is still in progress. In developing waste acceptance requirements, the PTB fulfills one of its duties as the institute responsible for waste disposal and gives guidelines for waste conditioning to waste producers and conditioners. (orig.) [de

Earthquakes - a danger to deep-lying repositories?; erdbeben: eine gefahr fuer tiefenlager?

Energy Technology Data Exchange (ETDEWEB)

NONE

2012-03-15

This booklet issued by the Swiss National Cooperative for the Disposal of Radioactive Waste NAGRA takes a look at geological factors concerning earthquakes and the safety of deep-lying repositories for nuclear waste. The geological processes involved in the occurrence of earthquakes are briefly looked at and the definitions for magnitude and intensity of earthquakes are discussed. Examples of damage caused by earthquakes are given. The earthquake situation in Switzerland is looked at and the effects of earthquakes on sub-surface structures and deep-lying repositories are discussed. Finally, the ideas proposed for deep-lying geological repositories for nuclear wastes are discussed.

Safety assessment for a KBS-3H spent nuclear fuel repository at Olkiluoto. Process report

Energy Technology Data Exchange (ETDEWEB)

Gribi, Peter; Johnson, Lawrence; Suter, Daniel; Smith, Paul; Pastina, Barbara; Snellman, Margit

2008-01-15

The KBS-3 method, based on multiple barriers, is the proposed spent fuel disposal method both in Sweden and Finland. KBS-3H and KBS-3V are the two design alternatives of the KBS-3 spent fuel disposal method. Posiva and SKB have conducted a joint research, demonstration and development (RDandD) programme in 2002-2007 with the overall aim of establishing whether KBS-3H represents a feasible alternative to the reference alternative KBS-3V. The overall objectives of the present phase covering the period 2004-2007 have been to demonstrate that the horizontal deposition alternative is technically feasible and to demonstrate that it fulfils the same long-term safety requirements as KBS-3V. The safety studies conducted as part of this programme include a safety assessment of a preliminary design of a KBS-3H repository for spent nuclear fuel located about 400 m underground at the Olkiluoto site, which is the proposed site for a spent fuel repository in Finland. In the KBS-3H design alternative, each canister, with a surrounding layer of bentonite clay, is placed in a perforated steel cylinder prior to emplacement; the entire assembly is called the supercontainer. Several supercontainers are positioned along parallel, 100-300 m long deposition drifts, which are sealed following waste emplacement using drift end plugs. Bentonite distance blocks separate the supercontainers, one from another, along the drift. Steel compartment plugs can be used to seal off drift sections with higher inflow, thus isolating the different compartments within the drift. The present report describes the main processes potentially affecting the long-term safety of the system, covering radiation-related, thermal, hydraulic, mechanical, chemical (including microbiological) and radionuclide transport-related processes. The process descriptions deal sequentially with the main sub-systems: fuel/cavity in canister, cast iron insert and copper canister, buffer and other bentonite components, supercontainer

Safety assessment for a KBS-3H spent nuclear fuel repository at Olkiluoto. Process report

International Nuclear Information System (INIS)

Gribi, Peter; Johnson, Lawrence; Suter, Daniel; Smith, Paul; Pastina, Barbara; Snellman, Margit

2008-01-01

The KBS-3 method, based on multiple barriers, is the proposed spent fuel disposal method both in Sweden and Finland. KBS-3H and KBS-3V are the two design alternatives of the KBS-3 spent fuel disposal method. Posiva and SKB have conducted a joint research, demonstration and development (RDandD) programme in 2002-2007 with the overall aim of establishing whether KBS-3H represents a feasible alternative to the reference alternative KBS-3V. The overall objectives of the present phase covering the period 2004-2007 have been to demonstrate that the horizontal deposition alternative is technically feasible and to demonstrate that it fulfils the same long-term safety requirements as KBS-3V. The safety studies conducted as part of this programme include a safety assessment of a preliminary design of a KBS-3H repository for spent nuclear fuel located about 400 m underground at the Olkiluoto site, which is the proposed site for a spent fuel repository in Finland. In the KBS-3H design alternative, each canister, with a surrounding layer of bentonite clay, is placed in a perforated steel cylinder prior to emplacement; the entire assembly is called the supercontainer. Several supercontainers are positioned along parallel, 100-300 m long deposition drifts, which are sealed following waste emplacement using drift end plugs. Bentonite distance blocks separate the supercontainers, one from another, along the drift. Steel compartment plugs can be used to seal off drift sections with higher inflow, thus isolating the different compartments within the drift. The present report describes the main processes potentially affecting the long-term safety of the system, covering radiation-related, thermal, hydraulic, mechanical, chemical (including microbiological) and radionuclide transport-related processes. The process descriptions deal sequentially with the main sub-systems: fuel/cavity in canister, cast iron insert and copper canister, buffer and other bentonite components, supercontainer

An approach for acquiring data for description of diffusion in safety assessment of radioactive waste repositories

International Nuclear Information System (INIS)

Vokal, A.; Vopalka, D.; Vecernik, P.; Institute of Chemical Technology in Prague, Prague

2010-01-01

Repositories for radioactive wastes are sited in the environment with very low permeability. One of the most important processes leading to the release of radionuclides to the environment is therefore diffusion of radionuclides in both natural and engineered barriers. Data for its description are crucial for the results of safety assessment of these repositories. They are obtained usually by comparison of the results of laboratory diffusion experiments with analytical and/or numerical solution of the diffusion equation with specified initial and boundary conditions. Results of the through-diffusion experiments are obviously evaluated by the 'time-lag' method that needs for most of sorbing species unfortunately very long time of the experiment duration. In this paper a modified approach is proposed for the evaluation of diffusion data for safety assessment, which decreases the influence of propagation uncertainties using incorrect data and reduces time for acquiring data for safety assessment. This approach consist in the following steps: (i) experimental measurement of material diffusion parameters under various conditions using non-sorbing tritiated water or chlorine for which it is easy to reach conditions under which the 'time-lag' method of evaluation of the result of the through-diffusion experiment is applicable - this step provides well established diffusion characteristics of materials for neutral species and anions, then (ii) to evaluate sorption isotherms for sorbing radionuclides from batch experiments under conditions corresponding to composition of material pore water, (iii) to assess the values of effective and apparent diffusion coefficients for sorbing radionuclides from well-defined diffusion coefficients of species in free water and (iv) to verify the obtained results using relatively short-term diffusion experiments with sorbing radionuclides, which will be evaluated using the time dependent decrease of the concentration in the input reservoir of

Site independent considerations on safety and protection of the groundwater - Basis for the fundamental evaluation of the licence granting for the surface buildings of a geological repository

International Nuclear Information System (INIS)

2013-08-01

This report explains how the protection of man and the environment can be assured for the surface facility of a deep geological repository. The report is intended primarily for the federal authorities, but also provides important information for the siting Cantons and siting regions. Nagra has also prepared an easily understandable brochure on the topic for the general public. The report was prepared at the request of the Swiss Federal Office of Energy (SFOE), with the aim of allowing the responsible federal authorities to evaluate, in a general manner, the aspects of safety and groundwater protection during the construction and operation of the surface facility of a geological repository, and the ability of the facility to fulfill the licensing requirements. The information is based on preliminary design concepts. The report presents the main features of a surface facility (design, activities), taking into account the waste to be emplaced in the repository and the potential conditions at the site. It is not a formal safety report for a facility at a real site within the context of licensing procedures as specified in the nuclear energy legislation. In line with the different legal and regulatory requirements, the following aspects are the subject of a qualitative analysis for the surface facility: (i) Nuclear safety and radiological protection during operation; (ii) Safety with respect to conventional (non-nuclear) accidents during operation and (iii) Protection of the groundwater during the construction and operational phases. The analysis highlights the fundamental requirements relating to the design of the surface facility, the operating procedures and the waste to be emplaced that have to be implemented in order to ensure the safety and protection of the groundwater. The influence of site-specific features and factors on the safety of the surface facility and on a possible impact on groundwater is also considered. To summarise, the report reaches the

Sorption data base for the cementitious near-field of L/ILW and ILW repositories for provisional safety analyses for SGT-E2

International Nuclear Information System (INIS)

Wieland, E.

2014-11-01

The near-field of the planned Swiss repositories for low- and intermediate-level waste (L/ILW) and long-lived intermediate-level waste (ILW) consists of large quantities of cementitious materials. Hardened cement paste (HCP) is considered to be the most important sorbing material present in the near-field of L/ILW and ILW repositories. Interaction of radionuclides with HCP represents the most important mechanism retarding their migration from the near-field into the host rock. This report describes a cement sorption data base (SDB) for the safety-relevant radionuclides in the waste that will be disposed of in the L/ILW and ILW repositories. The current update on sorption values for radionuclides should be read in conjunction with the earlier SDBs CEM-94, CEM-97 and CEM-02. Sorption values have been selected based on procedures reported in these earlier SDBs. The values are revised if corresponding new information and/or data are available. The basic information results from a survey of sorption studies published between 2002 and 2013. The sorption values recommended in this report have either been selected from in-house experimental studies or from literature data, and they were further assessed with a view to the sorption values recently published in the framework of the safety analysis for the planned near surface disposal facility in Belgium. The report summarizes the sorption properties of HCP and compiles sorption values for safety-relevant radionuclides and low-molecular weight organic molecules on undisturbed and degraded HCP. A list of the safety-relevant radionuclides is provided. The radionuclide inventories are determined by the waste streams to be disposed of in the L/ILW and ILW repositories. Information on the elemental and mineral composition of HCP was obtained from hydration studies. The concentrations of the most important impurity elements in cement were obtained from dissolution studies on HCP. Particular emphasis is placed on summarizing our

Modeling the impact of climate change in Germany with biosphere models for long-term safety assessment of nuclear waste repositories

International Nuclear Information System (INIS)

Staudt, C.; Semiochkina, N.; Kaiser, J.C.; Pröhl, G.

2013-01-01

Biosphere models are used to evaluate the exposure of populations to radionuclides from a deep geological repository. Since the time frame for assessments of long-time disposal safety is 1 million years, potential future climate changes need to be accounted for. Potential future climate conditions were defined for northern Germany according to model results from the BIOCLIM project. Nine present day reference climate regions were defined to cover those future climate conditions. A biosphere model was developed according to the BIOMASS methodology of the IAEA and model parameters were adjusted to the conditions at the reference climate regions. The model includes exposure pathways common to those reference climate regions in a stylized biosphere and relevant to the exposure of a hypothetical self-sustaining population at the site of potential radionuclide contamination from a deep geological repository. The end points of the model are Biosphere Dose Conversion factors (BDCF) for a range of radionuclides and scenarios normalized for a constant radionuclide concentration in near-surface groundwater. Model results suggest an increased exposure of in dry climate regions with a high impact of drinking water consumption rates and the amount of irrigation water used for agriculture. - Highlights: ► We model Biosphere Dose Conversion Factors for a representative group exposed to radionuclides from a waste repository. ► The BDCF are modeled for different soil types. ► One model is used for the assessment of the influence of climate change during the disposal time frame.

A Nuclide Release Model for a Deep Well Scenario near a Repository

International Nuclear Information System (INIS)

Lee, Youn Myoung; Jung, Jong Tae; Choi, Jong Won

2010-01-01

Recently several template programs ready for the safety assessment of a high-level radioactive waste repository (HLW) and a low- and intermediate level radioactive waste repository (LILW) systems, that are conceptually modeled, have been developed by utilizing GoldSim and AMBER at KAERI. During the last few years, such template programs have been utilized for the evaluation of nuclide transports in the nearand far-field of a repository as well as transport through the biosphere under various normal and disruptive release scenarios with assumed data. The GoldSim program, another template program by which influence due to a well located very near to the repository has been modeled and evaluated for an assumed case in order to simulate the worst exposure scenario. This seems very useful to evaluate an accidental event a long time after closure of the repository has been developed and illustrated

Retrievability as proposed in the US repository concept

International Nuclear Information System (INIS)

Harrington, P.G.

2000-01-01

The Nuclear Waste Policy Act states that any repository shall be designed and constructed to permit retrieval. Reasons for retrieval include public health and safety, environmental concerns, and recovery of economically valuable contents of spent nuclear fuel. The Nuclear Regulatory Commission requires that waste must be retrievable at any time up to 50 years after start of emplacement. The US Department of Energy intends to maintain a retrieval capability throughout the preclosure period. Possible preclosure periods range from a minimum of 50 years to as much as 300 years. Repository closure includes sealing all accessible portions of the repository, including ventilation shafts, access ramps and boreholes. Drip shields will be installed over the waste packages. Access to the repository after closure is not intended. The proposed repository includes horizontal emplacement drifts located in the unsaturated zone. The emplacement drift centerline spacing is 81 meters to provide a subboiling region between drifts for water drainage. A drip shield covers the waste packages. All emplacement drifts remain open until closure of the repository, providing performance benefits such as removing heat and moisture during the preclosure period and lowering postclosure temperatures. This does not impede retrieval, permitting a reversal of the emplacement process to accomplish retrieval under normal conditions. The preclosure period is therefore not to enhance retrievability, but does improve performance, and the resultant extension of the retrievability capability is a secondary effect. Information must be provided from the performance confirmation program to support a regulatory decision to close. Closure would isolate the repository from the accessible environment, preclude preferential flowpaths for water into the mountain, and minimize the possibility of inadvertent intrusion. (author)

Comparative analysis of safety related site characteristics

International Nuclear Information System (INIS)

Andersson, Johan

2010-12-01

This document presents a comparative analysis of site characteristics related to long-term safety for the two candidate sites for a final repository for spent nuclear fuel in Forsmark (municipality of Oesthammar) and in Laxemar (municipality of Oskarshamn) from the point of view of site selection. The analyses are based on the updated site descriptions of Forsmark /SKB 2008a/ and Laxemar /SKB 2009a/, together with associated updated repository layouts and designs /SKB 2008b and SKB 2009b/. The basis for the comparison is thus two equally and thoroughly assessed sites. However, the analyses presented here are focussed on differences between the sites rather than evaluating them in absolute terms. The document serves as a basis for the site selection, from the perspective of long-term safety, in SKB's application for a final repository. A full evaluation of safety is made for a repository at the selected site in the safety assessment SR-Site /SKB 2011/, referred to as SR-Site main report in the following

Comparative analysis of safety related site characteristics

Energy Technology Data Exchange (ETDEWEB)

Andersson, Johan (ed.)

2010-12-15

This document presents a comparative analysis of site characteristics related to long-term safety for the two candidate sites for a final repository for spent nuclear fuel in Forsmark (municipality of Oesthammar) and in Laxemar (municipality of Oskarshamn) from the point of view of site selection. The analyses are based on the updated site descriptions of Forsmark /SKB 2008a/ and Laxemar /SKB 2009a/, together with associated updated repository layouts and designs /SKB 2008b and SKB 2009b/. The basis for the comparison is thus two equally and thoroughly assessed sites. However, the analyses presented here are focussed on differences between the sites rather than evaluating them in absolute terms. The document serves as a basis for the site selection, from the perspective of long-term safety, in SKB's application for a final repository. A full evaluation of safety is made for a repository at the selected site in the safety assessment SR-Site /SKB 2011/, referred to as SR-Site main report in the following

The industrial organization of the repository. Pitfall or logical?

International Nuclear Information System (INIS)

Frostenson, Magnus

2010-11-01

From a systems perspective the organization of the Swedish final repository project for nuclear waste is studied. Different aspects of organization are identified in the report, covering dimensions of geographical, operative, structural, responsibility and contextual organization. Following SKB's site selection for the applications for the final repository for spent nuclear system and the closing of the surplus value agreement, issues concerning operative, structural and contextual organization tend to become particularly pressing, which is reflected in three research questions: - How will the final repository project be organized operatively and structurally over time? - Why is the final repository project organized in this way by SKB? - What kind of contextual organization takes place in the final repository project and what are the consequences of these activities? How the different industrial units of the final repository project should be run and within which structure, for example concerning ownership and integration of units, is established in the report. SKB's reasons for choosing this kind of organization are also highlighted. Apart from legal and safety-related demands that must be met together with the demands of the owners, SKB's strategic preference for insourcing conditions organizational choices. The traditional task centred operative and structural organization of SKB is also reflected in the organizational choices for the present and future units of the final depository system. Contextual organization implies deepened actor relationships between SKB's owners and SKB on the one side and the municipalities Oesthammar and Oskarshamn on the other. Through active organizing, the final repository arena 'narrows down' and the final repository issue turns into an in many respects local issue. There is a clear tendency that the roles of SKB are multiplied in order to handle the demands that central stakeholders - in particular the municipalities - place on

Using Specification and Description Language for Life Cycle Assesment in Buildings

Directory of Open Access Journals (Sweden)

Pau Fonseca i Casas

2017-06-01

Full Text Available The definition of a Life Cycle Assesment (LCA for a building or an urban area is a complex task due to the inherent complexity of all the elements that must be considered. Furthermore, a multidisciplinary approach is required due to the different sources of knowledge involved in this project. This multidisciplinary approach makes it necessary to use formal language to fully represent the complexity of the used models. In this paper, we explore the use of Specification and Description Language (SDL to represent the LCA of a building and residential area. We also introduce a tool that uses this idea to implement an optimization and simulation mechanism to define the optimal solution for the sustainability of a specific building or residential.

Site investigations for repositories for solid radioactive wastes in deep continental geological formations

International Nuclear Information System (INIS)

1982-01-01

This report reviews the earth-science investigations and associated scientific studies that may be needed to select a repository site and confirm that its characteristics are such that it will provide a safe confinement for solidified high-level and alpha-bearing and certain other solid radioactive wastes. Site investigations, as used in this report, cover earth sciences and associated safety analyses. Other site-investigation activities are identified but not otherwise considered here. The repositories under consideration are those consisting of mined cavities in deep continental rocks for accepting wastes in the solid and packaged form. The term deep as used in this report is used solely to emphasize the distinction between the repositories discussed in this report and those for shallow-ground disposal. In general, depths under consideration here are greater than 200 metres. The term continental refers to those geological formations that occur either beneath present-day land masses and adjoining islands or beneath the shallow seas. One of the objectives of site investigations is to collect the site-specific data necessary for the different evaluations, such as modelling required to assess the long-term safety of an underground repository

Site investigations for repositories for solid radioactive wastes in deep continental geological formations

Energy Technology Data Exchange (ETDEWEB)

1982-01-01

This report reviews the earth-science investigations and associated scientific studies that may be needed to select a repository site and confirm that its characteristics are such that it will provide a safe confinement for solidified high-level and alpha-bearing and certain other solid radioactive wastes. Site investigations, as used in this report, cover earth sciences and associated safety analyses. Other site-investigation activities are identified but not otherwise considered here. The repositories under consideration are those consisting of mined cavities in deep continental rocks for accepting wastes in the solid and packaged form. The term deep as used in this report is used solely to emphasize the distinction between the repositories discussed in this report and those for shallow-ground disposal. In general, depths under consideration here are greater than 200 metres. The term continental refers to those geological formations that occur either beneath present-day land masses and adjoining islands or beneath the shallow seas. One of the objectives of site investigations is to collect the site-specific data necessary for the different evaluations, such as modelling required to assess the long-term safety of an underground repository.

Study on a monitoring strategy to support decision making for geological repository closure

International Nuclear Information System (INIS)

Suyama, Yasuhiro; Tanabe, Hiromi; Eto, Jiro; Yoshimura, Kimitaka

2010-01-01

Japan currently plans to dispose of high-level radioactive wastes (vitrified HLWs) produced from the reprocessing of spent nuclear fuel in deep geological formations, in order to isolate the radioactive wastes from the human environment for tens of thousands of years. Such a geological repository must be designed to ensure operational safety and post-closure safety. Then, following the closure of the geological repository, post-closure safety will be provided by an engineered barrier system (EBS) and a natural barrier system (NBS) without relying on monitoring or institutional control. However, from a technical standpoint, monitoring has been required during backfilling in current studies. Additionally, there has been strong social pressure to continue monitoring during all the phases including post-closure. On the basis of the current situations, a monitoring strategy for geological disposal must be studied to ensure the long term safety of geological disposal. Focusing on decision making for geological repository closure, the authors have created a basic logical structure for the decision making process with the principles for ensuring safety and have developed a monitoring strategy based on the logical structure. The monitoring strategy is founded on three key aspects: the role of monitoring, boundary conditions of monitoring at the time of decision making, and a methodology for monitoring planning. Then, the monitoring strategy becomes a starting point of monitoring planning during site characterization, construction, operation and staged closure, as well as post-closure with institutional control, and of social science studies. (author)

Why is variability important for performance assessment and what are its consequences for site characterisation and repository design?

International Nuclear Information System (INIS)

Dverstorp, B.; Smith, P.A.; Zuidema, P.

1998-01-01

The importance of spatial variability is discussed in terms of its consequences for site characterisation and for repository design and safety. Variability is described in terms of various scales of discrete structural features and a pragmatic classification is proposed according to whether the features are: feasibility-determining (i.e. features within which repository construction and operation is not practical and which preclude long-term safety); layout-determining (i.e. features which, if avoided, would enhance long-term safety); safety-determining features (i.e. features which cannot be shown to be avoidable and which strongly influence the calculated long-term safety of the repository system). The significance with respect to the geosphere-transport barrier of small-scale pore structure within the various classes of feature is also discussed. The practical problems of characterising variability and modelling its effects on radionuclide transport are described. Key factors affecting groundwater flow and radionuclide transport are identified, models that incorporate spatial variability are described and the estimation of appropriate parameters for these models is discussed. (author)

Twenty years' experience with shallow ground repositories in India

International Nuclear Information System (INIS)

Balu, K.; Mohan, A.L.; Narayan, P.K.; Godse, V.B.; Sunder Rajan, N.S.

1984-01-01

With the setting up of the Bhabha Atomic Research Centre at Trombay, the first nuclear research centre in India, more than two decades ago, the need arose for disposal of the country's low-level radioactive solid wastes. Since then, nuclear power stations and other nuclear installations have been set up in different locations and the major mode of disposal for low-level solid wastes has been in engineered facilities in shallow land. The paper presents an overview of the Indian experience, with shallow land repositories for low-active solid wastes, from the point of view of design, operation and surveillance, at a number of sites. The influence of site characteristics on the design of the repository and the underground disposal modules is discussed. Details of the pre-operational investigations performed at different sites and the scheme for monitoring and surveillance of operating repositories are also included. The paper also presents briefly the type of safety analysis being carried out to evaluate possible environmental impact due to the operation of the shallow land repositories. (author)

Reference Design Description for a Geologic Repository

International Nuclear Information System (INIS)

2000-01-01

and move it below ground, as well as safety monitoring and closure. Section 7 describes the systems (natural and engineered) that ensure continued safety after closure. Section 8 offers design options that may be adopted in the future, and Section 9 provides a summary statement on the repository

Compliance status summaries for federal and state statutory directives that apply to the Salt Repository Project at the Deaf Smith County Site, Texas

International Nuclear Information System (INIS)

1986-07-01

This document contains statutory summaries, checklists of compliance requirements, status summaries, and lists of information needs for the environmental and health and safety statutory directives at Federal and State levels that apply to the Salt Repository Project at the Deaf Smith County Site, Texas. Statutes that apply in general to any repository project but not specifically to the Deaf Smith are not included. The information herein supplements the Salt Repository Project Statutory Compliance Plan and the Salt Repository Project Permitting Management Plan by providing lengthy details on statutory directives, compliance requirements, information needs, and the overall status of the environmental and health and safety compliance program for the Salt Repository Project at the Deaf Smith County Site, Texas

Project Guarantee 1985. Radioactive wastes: Properties and allocation to final repository types

International Nuclear Information System (INIS)

Anon.

1985-01-01

An overview of waste-specific data, as input into constructional engineering studies and safety analyses of Project Guarantee, is presented which describes the activity inventory of the radioactive waste to be disposed of, classified according to origin, the quantitative spezifications of the waste, the concept of classifying waste into appropriate categories, grouping into major categories and distribution of these between the different repository types, and finally, control measures which ensure observance of the specifications of the waste to be disposed of. It is expedient, for conceptional considerations and for the operational phase of the repository, to split the waste up into several suitably specified waste categories according to the practical aspects of origin and conditioning. This can be done in such a way that the waste within a specific category is sufficiently homogeneous with regard to its radiological properties and chemical composition for the requirements of safety analysis. The present volume contains base-data for around 30 waste types. Two waste types are documented with more detailed data as an example of the practicability of the comprehensive waste characterisation contained in reference report NTB 84-47. It is shown that waste-specific data which go into safety analysis and constructional engineering project studies are available in an appropriate degree of detail. The method of distributing the waste between repositories with differing degrees of protection and procedures for controlling adherence to admission specifications are developed and documented. It can be ensured that no waste with an impermissibly high radiotoxicity level will later be emplaced in a repository for low- and intermediate-level waste

Waste inventory, waste characteristics and waste repositories in Japan

International Nuclear Information System (INIS)

Shimooka, K.

1997-01-01

There are two types of repositories for the low level radioactive wastes in Japan. One is a trench type repository only for concrete debris generated from the dismantling of the research reactor. According to the safety assurance system, Japan Atomic Energy Research Institute (JAERI) has disposed of the concrete debris arose from the dismantling of the Japan Power Demonstration Reactor (JPDR). The other type is the concreted pit with engineered barriers. Rokkasho Low Level Radioactive Waste Disposal Center has this type of repository mainly for the power plant wastes. Japan Nuclear Fuel Ltd. (JNFL) established by electric power companies is the operator of the LLW disposal project. JNFL began the storage operation in 1992 and buried approximately 60,000 drums there. Two hundred thousand drums of uniformly solidified, waste may be buried ultimately. 4 refs, 3 tabs

Monitoring a repository for high-level radioactive waste in Germany. Possibilities and limits

International Nuclear Information System (INIS)

Jobmann, M.; Haverkamp, B.; Eilers, G.

2011-01-01

Pursuant to the new BMU safety requirements of September 2010 imposed upon the final storage of radioactive waste generating heat, the operator of a repository in Germany must establish a monitoring program which furnishes relevant measured information during the operations phase and for a defined period of time after closure of the repository. Within the framework of a feasibility study, an assessment basis was established to show in what format information about the status of a closed repository mine could be obtained technically without impairing the safety of barriers, for instance, by cable ducts. As a conceptual design basis, processes and measured quantities relevant to monitoring were attributed to the components of the current safety demonstration concept. For one model variant, monitoring possibilities of these processes were shown on the basis of monitoring modules. Some first experiments are being carried out in European underground laboratories about the use of wireless transmission systems in the repository area. On the basis of those activities, experiments could also be designed in the German exploratory mine of Gorleben in order to examine to what extent information obtained by monitoring could be transmitted in a wireless mode in rock salt formations. As far as the autonomous supply of electricity to measurement systems is concerned, which must be guaranteed on a long-term basis, there is now a possibility of using thermoelectric isotope generators or betavoltaic batteries. (orig.)

Workshop on Potentially Disruptive Phenomena for Nuclear Waste Repositories, July 27-28, 1977

International Nuclear Information System (INIS)

Jacobson, J.J.

1977-01-01

The workshop on Potentially Disruptive Phenomena for Nuclear Waste Repositories brought together experts in the geosciences to identify and evaluate potentially disruptive events and processes and to contribute ideas on how to extrapolate data from the past into the next one million years. The analysis is to be used to model a repository in geologic media for long-term safety assessments of nuclear waste storage. The workshop included invited presentations on the following items: an overview of the Waste Isolation Safety Assessment Program (WISAP), simulation techniques, subjective probabilities and methodology of obtaining data, similar modeling efforts at Lawrence Livermore and Sandia Laboratories, and geologic processes or events

Review of the potential effects of alkaline plume migration from a cementitious repository for radioactive waste

International Nuclear Information System (INIS)

Savage, D.

1997-01-01

Extensive use of cement and concrete is envisaged in the construction of geological repositories for low and intermediate-level radioactive wastes, both for structural, and encapsulation and backfilling purposes. Saturation of these materials with groundwater may occur in the post-closure period of disposal, producing a hyperalkaline pore fluid with a pH in the range 10-13.5. These pore fluids have the potential to migrate from the repository according to local groundwater flow conditions and react chemically with the host rock. These chemical reactions may affect the rock's capacity to retard the migration of radionuclides released from the repository after the degradation of the waste packages. The effects of these chemical reactions on the behaviour of the repository rock as a barrier to waste migration need to be investigated for the purposes of assessing the safety of the repository design (so-called 'safety assessment' or 'performance assessment'). The objectives of the work reported here were to: identify those processes influencing radionuclide mobility in the geosphere which could be affected by plume migration; review literature relevant to alkali-rock reaction; contact organisations carrying out relevant research and summarise their current and future activities; and make recommendations how the effects of plume migration can be incorporated into models of repository performance assessment. (author)

Assessment of the safety reserve offered by a concrete buffer in case of a geological repository in clay

International Nuclear Information System (INIS)

Govaerts, Joan; Weetjens, Eef; Marivoet, Jan

2012-01-01

Performance assessment calculations have been performed to investigate if the sorption of 14 C, 36 Cl and 129 I on the cementitious materials occurring in the near field of the repository on the diffusion would offer an extra safety reserve to deep disposal of vitrified HLW. Four cases have been studied: a reference case with no cementitious material and three cases in which the considered concrete region was subsequently extended to the buffer, backfill and gallery liner. The results show a beneficial impact on peak dose and residence time of the three radionuclides. The effect on total released fractions is very high for 14 C, moderate for 36 Cl and small for 129 I

Applicability of risk-informed criticality methodology to spent fuel repositories

International Nuclear Information System (INIS)

Mays, C.; Thomas, D.A.; Favet, D.

2000-01-01

An important objective of geologic disposal is keeping the fissionable material in a condition so that a self-sustaining nuclear chain reaction (criticality) is highly unlikely. This objective supports the overall performance objective of any repository, which is to protect the health and safety of the public by limiting radiological exposure. This paper describes a risk-informed, performance-based methodology, which combines deterministic and probabilistic approaches for evaluating the criticality potential of high-level waste and spent nuclear fuel after the repository is sealed and permanently closed (postclosure). (authors)

Performance analysis for waste repositories in the nordic countries. Report for project AFA-1.2

Energy Technology Data Exchange (ETDEWEB)

Vuori, S. [VTT Energy (Finland); Broden, K. [Studsvik RadWaste AB (Sweden); Carugati, S.; Brodersen, K. [Forskningscenter Risoe (Denmark); Walderhaug, T. [Icelandic Radiation Protection Institute (Iceland); Helgason, J. [Ekra Geological Consulting (Iceland); Sneve, M.; Hornkjoel, S. [Norwegian Radiation Protection (Norway); Backe, S. [IFE (Norway)

1997-02-01

The Nordic Nuclear Safety Research (NKS) project (AFA-1) focused on safety in the final disposal of long-lived low and medium level radioactive waste and its sub project (AFA-1.2), where this report has been produced, is dealing with the performance analysis of the engineered barrier system (near-field) of the repositories for low-and medium level wastes. The topic intentionally excludes the discussion of the characteristics of the geological host medium. Therefore a more generic discussion of the features of performance analysis is possible independent of the fact that different host media are considered in the Nordic countries. The different waste management systems existing and planned in the Nordic countries are shortly described in the report. In the report main emphasis is paid on the general repositories. Some of the phenomena and interactions relevant for a generic type of repository are discussed as well. Among the different approaches for the development of scenarios for safety and performance analyses one particular method - the Rock Engineering System (RES) - was chosen to be demonstratively tested in a brainstorming session, where the possible interactions and their safety significance were discussed employing a simplified and generic Nordic repository system as the reference system. As an overall impression, the AFA-project group concludes that the use of the RES approach is very easy to learn even during a short discussion session. The use of different ways to indicate the safety significance of various interactions in a graphical user interface increases the clarity. Within the project a simple software application was developed employing a generally available spread sheet programme. The developed tool allows an easy opportunity to link the cell specific comments readily available for the `reader` of the obtained results. A short review of the performance analyses carried out in the Nordic countries for actual projects concerning repositories for

Performance analysis for waste repositories in the nordic countries. Report for project AFA-1.2

International Nuclear Information System (INIS)

Vuori, S.; Broden, K.; Carugati, S.; Brodersen, K.; Walderhaug, T.; Helgason, J.; Sneve, M.; Hornkjoel, S.; Backe, S.

1997-02-01

The Nordic Nuclear Safety Research (NKS) project (AFA-1) focused on safety in the final disposal of long-lived low and medium level radioactive waste and its sub project (AFA-1.2), where this report has been produced, is dealing with the performance analysis of the engineered barrier system (near-field) of the repositories for low-and medium level wastes. The topic intentionally excludes the discussion of the characteristics of the geological host medium. Therefore a more generic discussion of the features of performance analysis is possible independent of the fact that different host media are considered in the Nordic countries. The different waste management systems existing and planned in the Nordic countries are shortly described in the report. In the report main emphasis is paid on the general repositories. Some of the phenomena and interactions relevant for a generic type of repository are discussed as well. Among the different approaches for the development of scenarios for safety and performance analyses one particular method - the Rock Engineering System (RES) - was chosen to be demonstratively tested in a brainstorming session, where the possible interactions and their safety significance were discussed employing a simplified and generic Nordic repository system as the reference system. As an overall impression, the AFA-project group concludes that the use of the RES approach is very easy to learn even during a short discussion session. The use of different ways to indicate the safety significance of various interactions in a graphical user interface increases the clarity. Within the project a simple software application was developed employing a generally available spread sheet programme. The developed tool allows an easy opportunity to link the cell specific comments readily available for the 'reader' of the obtained results. A short review of the performance analyses carried out in the Nordic countries for actual projects concerning repositories for

Deep geologic repository for low and intermediate radioactive level waste in Canada

International Nuclear Information System (INIS)

Liu Jianqin; Li Honghui; Sun Qinghong; Yang Zhongtian

2012-01-01

Ontario Power Generation (OPG) is undergoing a project for the long-term management of low and intermediate level waste (LILW)-a deep geologic repository (DGR) project for low and intermediate level waste. The waste source term disposed, geologic setting, repository layout and operation, and safety assessment are discussed. It is expected to provide reference for disposal of low and intermediate level waste that contain the higher concentration of long-lived radionuclides in China. (authors)

Geoscientific programme for investigation and evaluation of sites for the deep repository

International Nuclear Information System (INIS)

2000-08-01

SKB's goal is to be able to commence site investigations in 2002. Extensive preparations are now being made for this transition to the next phase in the siting process for the deep repository. This report describes the investigation and evaluation programme, which is focused on a deep repository in accordance with the KBS-3 method for spent nuclear fuel. When areas for site investigations have been chosen, the programmes will be adapted to the site-specific conditions. The geoscientific work during the site investigation phase is supposed to provide the broad knowledge base that is required to evaluate the suitability of investigated sites for a deep repository. The material must be comprehensive enough to: show whether the selected site satisfies fundamental safety requirements; permit comparisons with other investigated sites; and serve as a basis for adaptation of the deep repository to the properties and characteristics of the site with an acceptable impact on society and the environment. The work is being carried out in consultation with municipalities, regulatory authorities and nearby residents. The main product of the investigations is a site description, which presents collected data and interpreted parameters that are of importance both for the overall scientific understanding of the site and for the analyses and assessments that are made of design and safety assessment with respect to the deep repository's layout and construction as well as its long-term performance and radiological safety. The site description should furthermore present an integrated description of the site (geosphere and biosphere) and its regional environs with respect to current state and naturally ongoing processes. A technical risk evaluation is carried out, by which is meant a description of uncertainties in calculations and the environmental impact of the civil engineering work. The main product of the safety assessment is a safety report, which analyzes whether long-term safety

Geoscientific programme for investigation and evaluation of sites for the deep repository

Energy Technology Data Exchange (ETDEWEB)

NONE

2000-08-01

SKB's goal is to be able to commence site investigations in 2002. Extensive preparations are now being made for this transition to the next phase in the siting process for the deep repository. This report describes the investigation and evaluation programme, which is focused on a deep repository in accordance with the KBS-3 method for spent nuclear fuel. When areas for site investigations have been chosen, the programmes will be adapted to the site-specific conditions. The geoscientific work during the site investigation phase is supposed to provide the broad knowledge base that is required to evaluate the suitability of investigated sites for a deep repository. The material must be comprehensive enough to: show whether the selected site satisfies fundamental safety requirements; permit comparisons with other investigated sites; and serve as a basis for adaptation of the deep repository to the properties and characteristics of the site with an acceptable impact on society and the environment. The work is being carried out in consultation with municipalities, regulatory authorities and nearby residents. The main product of the investigations is a site description, which presents collected data and interpreted parameters that are of importance both for the overall scientific understanding of the site and for the analyses and assessments that are made of design and safety assessment with respect to the deep repository's layout and construction as well as its long-term performance and radiological safety. The site description should furthermore present an integrated description of the site (geosphere and biosphere) and its regional environs with respect to current state and naturally ongoing processes. A technical risk evaluation is carried out, by which is meant a description of uncertainties in calculations and the environmental impact of the civil engineering work. The main product of the safety assessment is a safety report, which analyzes whether long

Laymen's demand on an understandable safety analysis for a nuclear waste repository. A communication challenge

Energy Technology Data Exchange (ETDEWEB)

Andersson, T L; Thunberg, A M [KASAM - Swedish National Council for Nuclear Waste (Sweden)

1999-12-01

This paper is a summary in English of some impressions from a seminar 'Safety Analysis of the Final Disposal of Nuclear Waste. An issue for specialists only or for all of us?' The seminar was held in Swedish and was arranged by KASAM in Nykoeping, Sweden in November 1997. A report in Swedish from the seminar has been published. The seminar was arranged in response to a request from representatives from some of the municipalities concerned by the feasibility studies, which are part of the siting process. They had noticed that it is very hard for people without specialist competence to get an understanding of the safety issues based on the available information. There is a need for a presentation of the safety analysis, which is adopted not only for the need of the safety authorities, which have their own expertise, but also for the need of laymen who are involved in issues about the design, siting and safety of a final repository. Therefore, the seminar was mainly intended for representatives for the citizens (i.e. politicians) from the municipalities involved in the ongoing feasibility studies in Sweden. Some representatives from different environmental organisations opposing final disposal were also invited as well as representatives from the nuclear industry and from the concerned Swedish authorities. The seminar was structured in four sessions The handling of risk in the modern society - risk assessment and risk comparisons; The safety analysis and its role for the citizens; What can actually happen - in our own time and in the future?; Group discussions. In order to give a realistic picture of the intense debate, which at least in some of the municipalities had been very apparent, the organisers had chosen to make SKB and Greenpeace main actors at the seminar, such as they appeared in connection with campaign before the referendum at Malaa. Parts of the seminar were arranged like a hearing, led by a science journalist. The intention with this paper is not to

Laymen's demand on an understandable safety analysis for a nuclear waste repository. A communication challenge

International Nuclear Information System (INIS)

Andersson, T.L.; Thunberg, A.M.

1999-01-01

This paper is a summary in English of some impressions from a seminar 'Safety Analysis of the Final Disposal of Nuclear Waste. An issue for specialists only or for all of us?' The seminar was held in Swedish and was arranged by KASAM in Nykoeping, Sweden in November 1997. A report in Swedish from the seminar has been published. The seminar was arranged in response to a request from representatives from some of the municipalities concerned by the feasibility studies, which are part of the siting process. They had noticed that it is very hard for people without specialist competence to get an understanding of the safety issues based on the available information. There is a need for a presentation of the safety analysis, which is adopted not only for the need of the safety authorities, which have their own expertise, but also for the need of laymen who are involved in issues about the design, siting and safety of a final repository. Therefore, the seminar was mainly intended for representatives for the citizens (i.e. politicians) from the municipalities involved in the ongoing feasibility studies in Sweden. Some representatives from different environmental organisations opposing final disposal were also invited as well as representatives from the nuclear industry and from the concerned Swedish authorities. The seminar was structured in four sessions The handling of risk in the modern society - risk assessment and risk comparisons; The safety analysis and its role for the citizens; What can actually happen - in our own time and in the future?; Group discussions. In order to give a realistic picture of the intense debate, which at least in some of the municipalities had been very apparent, the organisers had chosen to make SKB and Greenpeace main actors at the seminar, such as they appeared in connection with campaign before the referendum at Malaa. Parts of the seminar were arranged like a hearing, led by a science journalist. The intention with this paper is not to

Single or multi-repository concept? A personal contribution to the discussion of pros and cons

International Nuclear Information System (INIS)

Closs, K.-D.; Duphorn, K.; Kuehn, K.

2002-01-01

The coalition agreement concluded between the parties in government, i.e. SPD and Alliance 90/The Greens, after the 1998 federal elections provided for an approach to back-end fuel cycle conditions for German nuclear power plants which differed from the practice intended up to that point. Among other things, only one single repository in deep geologic formations is to be available approximately from 2030 on. Exploration of the Gorleben salt dome is to be interrupted because of prevailing doubts while, at the same time, a basic discussion is to be held about potentially suitable host rock varieties and sites. The 'Working Party of Selection Procedures for Repository Sites (AkEnd)' instituted by the federal government is to elaborate scientific siting criteria under these conditions. The authors, who are members of AkEnd, come to the conclusion, based on technical findings and the historical back-end fuel cycle and siting situation in Germany, that a multi-repository concept incorporates substantial advantages. These are the aspects referred to: - Back-end fuel cycle concept: The multi-repository concept offers clearcut advantages in the far advanced Konrad repository project. This is true also against the background of the volume of waste with negligible heat generation to be managed by public authorities. - Long-term safety: Advantages exist, among other things, with respect to technical safety regarding possible compromises in a single-repository concept and the possible restriction of the number of potential sites. - Methods of detection: Spreading the waste over two or more repositories offers benefits in the technical aspects of detection. - Cost: A differentiated assessment of the cost arising in connection with the total amount of waste to be stored in repositories indicates benefits of a multi-repository concept, especially for the public authorities. (orig.) [de

Some reflections on human intrusion into a nuclear waste repository

International Nuclear Information System (INIS)

Westerlind, M.

2002-01-01

This paper summarises some of the Swedish nuclear regulators' requirements and views related to intrusion into a repository for spent nuclear fuel, in the post-closure phase. The focus is however on experiences from the interaction with various stakeholders in the Swedish process for siting a repository. It is recognised that intrusion is not a major concern but that it is regularly raised in the debate, often in connection with issues related to retrievability. It is pointed out that more attention should be paid to the repository performance after an intrusion event, both in safety assessments and in communication with stakeholders, and not only address the immediate impacts to intruders. It is believed that international co-operation would be useful for developing methodologies for defining intrusion scenarios. (author)

The role of natural analogues in safety assessment and acceptability

International Nuclear Information System (INIS)

Papp, Toenis

1987-01-01

The safety assessment must evaluate the level of safety for a repository, the confidence that can be placed on the assessment and how well the repository can meet the acceptance criteria of the society. Many of the processes and phenomena that govern the long term performance of a deep geologic repository for radioactive waste also take place in nature. To investigate these natural analogues and try to validate the models on which the safety assessment are based is a main task in the effort to build of confidence in the safety assessments. The assessment of the safety of a repository can, however, not only be based on good models. The possible role of natural analogues or natural evidence in other parts of the safety assessment is discussed. Specially with regard to - the need to demonstrate that all relevant processes have been taken into account, and that the important ones have been validated to an acceptable level for relevant parameters spans, -the definition and analysis of external scenarios for the safety assessment and for the claim that all reasonable scenarios have been addressed, - the public confidence in the long-term relevance of the acceptance criteria. (author)

Technical realization of a closure concept for a chamber-system in the underground Richard repository in the Czech Republic

International Nuclear Information System (INIS)

Kucerka, Miroslav

2006-01-01

The Phare project CZ 632.02.04 'Realization of closure of a chamber in the Richard repository as input for establishing a safety case' is a follow up implementation phase of the Phare project, CZ 01.14.03 'Solution for closure of a chamber in the Richard repository'. Main objective of both projects is to propose and realize a disposal system in selected chambers of the Richard repository, which will eliminate burden from the past practices in waste management during the first phase of the Richard repository operation (1965 - 1980) and which will improve its overall long term safety. This objective will be assured by realization of the concept of so called 'hydraulic cage', which technical solution was developed by DBE Technology within the Phare project CZ 01.14.03. The solution is described in the previous presentation 'Hydraulic Cage Concept for Waste Chambers and its Technical Implementation for the Underground Richard Repository, Litomerice, Czech Republic'. (author)

Assessments of the thermal evolution for a radioactive waste final repository using analytical methods

International Nuclear Information System (INIS)

Radut, A. C.; Roman, M. R.; Florea, S.; Ionescu, D. V.; Olteanu, G.; Valeca, S.

2016-01-01

In the designing process for a radioactive final repository concept, the temperature evolution over time is a significant issue for the stability and long-term safety of entire emplacement. In particular, the maximum value of temperature in the whole structure, during time, must not exceed a certain safety value which depends, beside other criteria, on the bulk material of the repository. A computer code TEMPROC, based on analytical model for the transient thermal heat conduction, described in this paper, was developed inside ''Fuel Performance''Department from ICN Pitesti, in order to evaluate the waste repository's temperature distribution. The program was developed under ''Microsoft FORTRAN Power Station''platform that provides IMSL subroutines library support for numeric algorithm. So the program is relative small, with a good calculus speed. The numerical results obtained with TEMPROC computer code, have been acceptably compared with similar existing data from scientific literature [1]. (authors)

Safety assessment input for site selection - the Swedish example - 59031

International Nuclear Information System (INIS)

Andersson, Johan

2012-01-01

Svensk Kaernbraenslehantering AB (SKB) has performed comprehensive investigations of two candidate sites for a final repository for Sweden's spent nuclear fuel. In March 2011 SKB decided to submit licence applications for a final repository at Forsmark. Before selection, SKB stated that the site that offers the best prospects for achieving long-term safety in practice would be selected. Based on experiences previous safety assessments, a number of issues related to long-term safety need to be considered in the context of site comparison. The factors include sensitivity to climate change such as periods of permafrost and glaciations, rock mechanics evolution including the potential for thermally induced spalling and sensitivity to potential future earthquakes, current and future groundwater flow, evolution of groundwater composition and proximity to mineral resources. Each of these factors related to long-term safety for the two candidate sites is assessed in a comparative analysis of site characteristics. The assessment also considers differences in biosphere conditions and in the confidence of the site descriptions. The comparison is concluded by an assessment on how the identified differences would affect the estimated radiological risk from a repository located at either of the sites. The assessment concludes that there are a number of safety related site characteristics for which the analyses do not show any decisive differences in terms of implications on safety, between the sites Forsmark and Laxemar. However, the frequency of water conducting fractures at repository depth is much smaller at Forsmark than at Laxemar. This difference, in turn, affects the future stability of the current favourable groundwater composition, which combined with the much higher flows at Laxemar would, for the current repository design, lead to a breach in the safety functions for the buffer and the canister for many more deposition positions at Laxemar than at Forsmark. Thereby

Preclosure radiological safety analysis for accident conditions of the potential Yucca Mountain Repository: Underground facilities; Yucca Mountain Site Characterization Project

Energy Technology Data Exchange (ETDEWEB)

Ma, C.W.; Sit, R.C.; Zavoshy, S.J.; Jardine, L.J. [Bechtel National, Inc., San Francisco, CA (United States); Laub, T.W. [Sandia National Labs., Albuquerque, NM (United States)

1992-06-01

This preliminary preclosure radiological safety analysis assesses the scenarios, probabilities, and potential radiological consequences associated with postulated accidents in the underground facility of the potential Yucca Mountain repository. The analysis follows a probabilistic-risk-assessment approach. Twenty-one event trees resulting in 129 accident scenarios are developed. Most of the scenarios have estimated annual probabilities ranging from 10{sup {minus}11}/yr to 10{sup {minus}5}/yr. The study identifies 33 scenarios that could result in offsite doses over 50 mrem and that have annual probabilities greater than 10{sup {minus}9}/yr. The largest offsite dose is calculated to be 220 mrem, which is less than the 500 mrem value used to define items important to safety in 10 CFR 60. The study does not address an estimate of uncertainties, therefore conclusions or decisions made as a result of this report should be made with caution.

Natural analogues for containment-providing barriers for a HLW repository in salt

Energy Technology Data Exchange (ETDEWEB)

Wolf, J.; Noseck, U.

2015-06-15

In 2005, a German research project was started to develop a novel approach to prove safety for a HLW repository in a salt formation, to refine the safety concept, to identify open scientific issues and to define necessary R&D work. This project aimed at identifying the key information for a HLW repository in salt. One important question is how this information may be best fulfilled by natural analogue studies. This question is answered by starting a review of the required key information needs of the safety case (post-closure phase) in order to assess whether or not these requirements can be supported by natural analogues information. In order to structure the review and to address the key elements of the safety concepts, three types of natural analogues are distinguished: (i) natural analogues for the integrity of the geological barrier, (ii) natural analogues for the integrity of the geotechnical barriers and (iii) natural analogues for release scenarios. For the safety case in salt type (i) and (ii) are of highest importance and are treated in this paper. The assessment documented in this paper on the one hand indicates the high potential benefit of natural analogues for a safety case in salt and on the other hand helps to focus the available human and financial resources for the safety case on the most safety-relevant aspects. (authors)

Safety case for the disposal of spent nuclear fuel at Olkiluoto - Synthesis 2012

International Nuclear Information System (INIS)

2012-12-01

TURVA-2012 is Posiva's safety case in support of the Preliminary Safety Analysis Report (PSAR 2012) and application for a construction licence for a spent nuclear fuel repository. Consistent with the Government Decisions-in- Principle, this foresees a repository developed in bedrock at the Olkiluoto site according to the KBS-3 method, designed to accept spent nuclear fuel from the lifetime operations of the Olkiluoto and Loviisa reactors. Synthesis 2012 presents a synthesis of Posiva Oy's Safety Case 'TURVA-2012' portfolio. It summarises the design basis for the repository at the Olkiluoto site, the assessment methodology and key results of performance and safety assessments. It brings together all the lines of argument for safety, evaluation of compliance with the regulatory requirements, and statement of confidence in long-term safety and Posiva's safety analyses. The TURVA-2012 safety case demonstrates that the proposed repository design provides a safe solution for the disposal of spent nuclear fuel, and that the performance and safety assessments are fully consistent with all the legal and regulatory requirements related to long-term safety as set out in Government Decree 736/2008 and in guidance from the nuclear regulator - the STUK. Moreover, Posiva considers that the level of confidence in the demonstration of safety is appropriate and sufficient to submit the construction licence application to the authorities. The assessment of long-term safety includes uncertainties, but these do not affect the basic conclusions on the long-term safety of the repository. (orig.)

SR-can: preliminary feedback to canister fabrication, repository design and future R and D

International Nuclear Information System (INIS)

Hedin, A.; Sellin, P.

2007-01-01

This paper discusses preliminary feedback from SKB's on-going safety assessment SR-Can, to he finalized in 2006. The assessment, which is not part of a formal licence application, is an important step towards the SR-Site assessment to be delivered in 2008 and which will support a licence application for a Swedish deep repository for spent nuclear fuel. The SR-Can assessment will use data from the initial stage of the on-going site investigations at the two candidate sites at Forsmark and Oskarshamn. Review comments on SR-Can from Swedish authorities are expected in the summer of 2007 and these will be taken into account when preparing the SR-Site assessment. An Interim version of the SR-Can report was produced in September 2004 and has been reviewed by the Swedish authorities supported by an international review team. The assessment concerns a KBS 3 repository for which the key safety related features can be summarised in the primary safety function isolation and the secondary function retardation. The isolation function is more prominent in the KBS 3 method compared to many other repository concepts. (authors)

Safety and ethical aspects on retrievability: A Swedish nuclear regulator's view

International Nuclear Information System (INIS)

Toverud, Oe.; Wingefors, S.

2000-01-01

An important contribution to the discussion on retrieval in Sweden has been the ethical principle of the Swedish National Council for Nuclear Waste (KASAM). ''The KASAM Principle'' means that the present generation, which has reaped the benefits of nuclear energy, must also take care of the waste and not transfer the responsibility to future generations; a repository should be designed and constructed so that monitoring and remedial actions are not necessary in the future. However, future generations, probably with better knowledge and other values, must still have the freedom to make their own decisions; we should therefore not make monitoring and remedial action unnecessarily difficult. SKI generally supports the KASAM principle but its application in the individual case should be based on solid evidence that both aspects have been covered in a suggested repository design. There may be a number of possible reasons for retrieval of spent nuclear fuel from a repository and they range from technical to purely political. SKI supports that the repository shall not be designed so that it unnecessarily impairs future attempts to retrieve the waste, monitor or ''repair'' the repository. However, measures to facilitate any kind of access to the repository must not reduce the long term safety of the repository. SKI concludes that: Future generations may wish to retrieve the spent fuel from a sealed repository. Disposal method and repository design should consider this and not make such retrieval unnecessarily difficult. On the other hand, any measures taken to facilitate retrieval must not significantly impair the long term safety functions of the repository. It must be shown that the safety aspects have been adequately considered. Retrievability must always be discussed with caution, so that it will not give the impression of doubts concerning the safety of the repository. (author)

Improvement of storage conditions and closure of the radioactive waste repository - Rozan

International Nuclear Information System (INIS)

Dutton, L.M.C.; Pacey, N.R.; Buckley, M.J.; Thomson, J.G.; Miller, W.; Barraclough, I.; Tomczak, W.; Mitrega, J.; Smietanski, L.

2005-01-01

The Rozan repository is a near-surface repository on the site of an ex-military fort, operated by Radioactive Waste Management Plant (RWMP). Solid or encapsulated waste is consigned to the repository. Low- and medium-activity waste produced in Poland is collected, processed, solidified and prepared for disposal at the Swierk facility. The waste is currently stored or disposed of within the fort structures, these have robust concrete walls, that provide both shielding and containment. The project, funded by the European Commission through the Phare Programme, aimed to improve the storage conditions and determine a strategy for closure achieving two key results; Stakeholder agreement to a strategy for the management and closure of the repository, and; Approval by the National Atomic Energy Agency of the safety case for the selected strategy. The strategy was selected using a multi-criteria analysis methodology at workshops that involved experts, regulators and other stakeholders. The selected strategy proposed that the waste in Facilities 3A and 8 should be left in situ and these facilities should continue to operate until the repository is closed. The waste in Rooms K7 to K9 of Facility 1 and in Facilities 2 and 3 should be retrieved, assayed, treated and packaged prior to redisposal. The short-lived waste that is retrieved from Rooms K7 to K9 of Facility 1 and Facilities 2 and 3 should be emplaced in Facility 8 subject to acceptance by the NAEA of the dose of 0.3mSv/y that might occur at long times in the future from a very unlikely scenario. When operations at the repository end, Facilities 3A and 8 should be covered with a multi-layer cap. Following selection of the strategy, assessment work was undertaken to support the production of the suite of safety cases. (author)

Nuclide release calculation in the near-field of a reference HLW repository

International Nuclear Information System (INIS)

Lee, Youn Myoung; Hwang, Yong Soo; Kang, Chul Hyung

2004-01-01

The HLW-relevant R and D program for disposal of high-level radioactive waste has been carried out at Korea Atomic Energy Research Institute (KAERI) since early 1997 in order to develop a conceptual Korea Reference Repository System for direct disposal of nuclear spent fuel by the end of 2007. A preliminary reference geologic repository concept considering such established criteria and requirements as waste and generic site characteristics in Korea was roughly envisaged in 2003 focusing on the near-field components of the repository system. According to above basic repository concept, which is similar to that of Swedish KBS-3 repository, the spent fuel is first encapsulated in corrosion resistant canisters, even though the material has not yet been determined, and then emplaced into the deposition holes surrounded by high density bentonite clay in tunnels constructed at a depth of about 500 m in a stable plutonic rock body. Not only to demonstrate how much a reference repository is safe in the generic point of view with several possible scenarios and cases associated with a preliminary repository concept by conducting calculations for nuclide release and transport in the near-field components of the repository, even though enough information has not been available that much yet, but also to show a methodology by which a generic safety assessment could be performed for further development of Korea reference repository concept, nuclide release calculation study strongly seems to be necessary

Handling of future human actions in the safety assessment SR-Site

International Nuclear Information System (INIS)

2010-12-01

This report documents the future human actions, FHA, considered in the long-term safety analysis of a KBS-3 repository. The report is one of the supporting documents to the safety assessment SR-Site (see further the Main report /SKB 2011/). The purpose of this report is to provide an account of general considerations concerning FHA, the methodology applied in SR-Site to assess FHA, the aspects of FHA needed to be considered in the evaluation of their impact on a deep geological repository and to select and analyse representative scenarios for illustrative consequence analysis. The main focus of this report is a time period when institutional control has ceased to be effective, thereby permitting inadvertent intrusion. However, a brief discussion of the earlier period when the repository has been closed, sealed and continuously kept under institutional control is also provided. General The potential exposure to large quantities of radiotoxic material is an inescapable consequence of the deposition of spent nuclear fuel in a final repository, and consequently intrusion into the repository needs to be considered in repository design and safety assessment. In accordance with ICRP recommendations /ICRP 2000/, intrusion in the post-closure phase of institutional control and beyond is primarily prevented through the design of the repository. In addition to that there will presumably continue to be safeguards measures, preservation of information (record keeping) and possibly some sort of markers placed at the site. During the institutional control period, activities at the site have to be restricted or directed if they have the potential to interfere with or hinder surveillance of the site, but this does not necessarily rule out all forms of access to the area. Also the fact that the repository contains fissile materials is an important aspect. Control of safeguards measures will most likely be upheld by national as well as international agencies. Furthermore, the

Handling of future human actions in the safety assessment SR-Site

Energy Technology Data Exchange (ETDEWEB)

2010-12-15

This report documents the future human actions, FHA, considered in the long-term safety analysis of a KBS-3 repository. The report is one of the supporting documents to the safety assessment SR-Site (see further the Main report /SKB 2011/). The purpose of this report is to provide an account of general considerations concerning FHA, the methodology applied in SR-Site to assess FHA, the aspects of FHA needed to be considered in the evaluation of their impact on a deep geological repository and to select and analyse representative scenarios for illustrative consequence analysis. The main focus of this report is a time period when institutional control has ceased to be effective, thereby permitting inadvertent intrusion. However, a brief discussion of the earlier period when the repository has been closed, sealed and continuously kept under institutional control is also provided. General The potential exposure to large quantities of radiotoxic material is an inescapable consequence of the deposition of spent nuclear fuel in a final repository, and consequently intrusion into the repository needs to be considered in repository design and safety assessment. In accordance with ICRP recommendations /ICRP 2000/, intrusion in the post-closure phase of institutional control and beyond is primarily prevented through the design of the repository. In addition to that there will presumably continue to be safeguards measures, preservation of information (record keeping) and possibly some sort of markers placed at the site. During the institutional control period, activities at the site have to be restricted or directed if they have the potential to interfere with or hinder surveillance of the site, but this does not necessarily rule out all forms of access to the area. Also the fact that the repository contains fissile materials is an important aspect. Control of safeguards measures will most likely be upheld by national as well as international agencies. Furthermore, the

SR 97 - Waste, repository design and sites. Background report to SR 97 SKB

International Nuclear Information System (INIS)

1999-10-01

SR 97 is a comprehensive analysis of long-term safety of a deep repository for spent nuclear fuel. The repository is assumed to be designed according to the KBS-3 method. Assessments are performed in SR 97 for three fictitious sites: Aberg, Beberg and Ceberg. One premise is that data used for assessment of the fictitious sites are to be taken from sites that have previously been investigated. The spent nuclear fuel is enclosed in copper canisters with an insert of cast iron. The canisters are emplaced in bored holes in the floor of the deposition tunnels. Around each canister, bentonite blocks are stacked which, after absorbing water and swelling, will isolate the canister from groundwater, hold the canister in place and retard transport of radionuclides from the canister to the surrounding rock. The spent nuclear fuel will emit heat for a long time, due to the decay heat. The maximum permissible temperature on the canister surface has been chosen at 100 deg C. The spacing between the deposition holes and between the deposition tunnels is adjusted site-specifically to meet this requirement. The thermal properties of the rock and the buffer material are of importance for how closely the deposition holes and tunnels can be spaced. After deposition, the deposition tunnels are backfilled with a mixture of bentonite and crushed rock. SR 97 examines above all the consequences of various scenarios and the handling of various types of uncertainties. The different repository sites illustrate normal properties for Swedish bedrock which are of importance for safety. To facilitate the work, the repositories on the three sites are configured as similarly as possible, which means for example that they are located at roughly the same depth and are fitted into the bedrock in a relatively similar fashion. Apart from the siting of a repository for spent nuclear fuel, the site may need to house a separate repository for other long-lived waste. This possibility has been considered in

SR 97 - Waste, repository design and sites. Background report to SR 97 SKB

Energy Technology Data Exchange (ETDEWEB)

NONE

1999-10-01

SR 97 is a comprehensive analysis of long-term safety of a deep repository for spent nuclear fuel. The repository is assumed to be designed according to the KBS-3 method. Assessments are performed in SR 97 for three fictitious sites: Aberg, Beberg and Ceberg. One premise is that data used for assessment of the fictitious sites are to be taken from sites that have previously been investigated. The spent nuclear fuel is enclosed in copper canisters with an insert of cast iron. The canisters are emplaced in bored holes in the floor of the deposition tunnels. Around each canister, bentonite blocks are stacked which, after absorbing water and swelling, will isolate the canister from groundwater, hold the canister in place and retard transport of radionuclides from the canister to the surrounding rock. The spent nuclear fuel will emit heat fora long time, due to the decay heat. The maximum permissible temperature on the canister surface has been chosen at 100 deg C. The spacing between the deposition holes and between the deposition tunnels is adjusted site-specifically to meet this requirement. The thermal properties of the rock and the buffer material are of importance for how closely the deposition holes and tunnels can be spaced. After deposition, the deposition tunnels are backfilled with a mixture of bentonite and crushed rock. SR 97 examines above all the consequences of various scenarios and the handling of various types of uncertainties. The different repository sites illustrate normal properties for Swedish bedrock which are of importance for safety. To facilitate the work, the repositories on the three sites are configured as similarly as possible, which means for example that they are located at roughly the same depth and are fitted into the bedrock in a relatively similar fashion. Apart from the siting of a repository for spent nuclear fuel, the site may need to house a separate repository for other long-lived waste. This possibility has been considered in

Conceptualization and software development of a simulation environment for probalistic safety assessment of radioactive waste repositories

Energy Technology Data Exchange (ETDEWEB)

Ghofrani, Javad

2016-05-26

Uncertainty and sensitivity analysis of complex simulation models are prominent issues, both in scientific research and education. ReSUS (Repository Simulation, Uncertainty propagation and Sensitivity analysis) is an integrated platform to perform such analysis with numerical models that simulate the THMC (Thermal Hydraulical Mechanical and Chemical) coupled processes via different programs, in particular in the context of safety assessments for radioactive waste repositories. This thesis presents the idea behind the software platform ReSUS and its working mechanisms. Apart from the idea and the working mechanisms, the thesis describes applications related to the safety assessment of radioactive waste disposal systems. In this thesis, previous simulation tools (including the preceding version of ReSUS) are analyzed in order to provide a comprehensive view of the state of the art. In comparison to this state, a more sophisticated software tool is developed here, which provides features which are not offered by previous simulation tools. To achieve this objective, the software platform ReSUS provides a framework for handling probabilistic data uncertainties using deterministic external simulation tools, thus enhancing uncertainty and sensitivity analysis. This platform performs probabilistic simulations of various models, in particular THMC coupled processes, using stand-alone deterministic simulation software tools. The complete software development process of the ReSUS Platform is discussed in this thesis. ReSUS components are developed as libraries, which are capable of being linked to other code implementations. In addition, ASCII template files are used as means for uncertainty propagation into the input files of deterministic simulation tools. The embedded input sampler and analysis tools allow for sensitivity analysis in several kinds of simulation designs. The novelty of the ReSUS platform consists in the flexibility to assign external stand-alone software

Conceptualization and software development of a simulation environment for probalistic safety assessment of radioactive waste repositories

International Nuclear Information System (INIS)

Ghofrani, Javad

2016-01-01

Uncertainty and sensitivity analysis of complex simulation models are prominent issues, both in scientific research and education. ReSUS (Repository Simulation, Uncertainty propagation and Sensitivity analysis) is an integrated platform to perform such analysis with numerical models that simulate the THMC (Thermal Hydraulical Mechanical and Chemical) coupled processes via different programs, in particular in the context of safety assessments for radioactive waste repositories. This thesis presents the idea behind the software platform ReSUS and its working mechanisms. Apart from the idea and the working mechanisms, the thesis describes applications related to the safety assessment of radioactive waste disposal systems. In this thesis, previous simulation tools (including the preceding version of ReSUS) are analyzed in order to provide a comprehensive view of the state of the art. In comparison to this state, a more sophisticated software tool is developed here, which provides features which are not offered by previous simulation tools. To achieve this objective, the software platform ReSUS provides a framework for handling probabilistic data uncertainties using deterministic external simulation tools, thus enhancing uncertainty and sensitivity analysis. This platform performs probabilistic simulations of various models, in particular THMC coupled processes, using stand-alone deterministic simulation software tools. The complete software development process of the ReSUS Platform is discussed in this thesis. ReSUS components are developed as libraries, which are capable of being linked to other code implementations. In addition, ASCII template files are used as means for uncertainty propagation into the input files of deterministic simulation tools. The embedded input sampler and analysis tools allow for sensitivity analysis in several kinds of simulation designs. The novelty of the ReSUS platform consists in the flexibility to assign external stand-alone software

Preclosure safety analysis for a prospective Yucca Mountain conceptual design repository

International Nuclear Information System (INIS)

Ma, C.W.; Jardine, L.J.

1989-12-01

A preliminary probabilistic risk assessment was performed for the prospective Yucca Mountain conceptual design repository. A new methodology to quantify radioactive source terms was developed and applied in the analysis. The study identified 42 event trees comprising 278 accident scenarios. The maximum offsite dose evaluated in this study is about 1000 mrem. For the majority of the accident scenarios, either the offsite dose is less than 100 mrem or the probability of occurrence is less than 1 x 10 -9 /yr. Only 11 accident scenarios with a dose larger than 100 mrem and an associated probability greater than 1 x 10 -9 /yr were identified. A more detailed follow-on analysis for seismic events of various severity was also performed, and similar results were obtained. Therefore, based on the results of this analysis, no significant risk to the general public was identified during the preclosure period for the conceptual repository design. 13 refs., 4 figs., 2 tabs

Safety analysis of the post-operational phase

International Nuclear Information System (INIS)

Berg, H.P.; Ehrlich, D.

1991-01-01

The safety analysis of normal operation covers an analytical study of the system parts ultimate repository - waste forms of the ultimate repository system under normal and accidental operation. On that basis a requirement concept has been developed which entails reactions on planning and design of the repository, and requirements of waste products, packagings and permissible activities. The procedure for the operational phase is explained giving the Konrad repository project as an example. (DG) [de

Use of modeling in repository licensing

International Nuclear Information System (INIS)

McGarry, J.M. III; Echols, F.S.

1995-01-01

A review of the regulatory history of the Nuclear Regulatory Commission (NRC) regulations applicable to the licensing of a geologic repository, as well as a review of NRC administrative (licensing) decisions and federal case law, support the NRC's use of simplified models, in appropriate circumstances, which provide well-documented and reasonably conservative bounding assumptions, together with the use of expert judgement, natural analogues, and other aids to supplement available information, in reaching its reasonable assurance determination whether the public health and safety will be adequately protected if the Yucca Mountain, Nevada site should be licensed for development as a geologic repository. Specific examples are provided to assist the reader to better understand how such qualitative concepts as open-quote reasonable assurance close-quote, open-quote reasonably conservative close-quote, and open-quote adequate close-quote protection are used in an administrative context to resolve technical issues

Final repositories for high-level radioactive waste; Endlagerung hochradioaktiver Abfaelle

Energy Technology Data Exchange (ETDEWEB)

NONE

2015-10-15

The brochure on final repositories for high-level radioactive waste covers the following issues: What is the origin of radioactive wastes? How large are the waste amounts? What is going to happen with the wastes? What is the solution for the Waste disposal? A new site search is started. Safety requirements for the final disposal of high-level radioactive wastes. Comparison of host rocks. Who is responsible and who will pay? Final disposal of high-level radioactive wastes worldwide. Short summary: History of the search for a final repository for high-level radioactive wastes in Germany.

Identification and applicability of analogues for a safety case for a HLW repository in evaporites: results from a NEA workshop

Energy Technology Data Exchange (ETDEWEB)

Noseck, U.; Wolf, J. [Gesellschaft für Anlagen und Reaktorsicherheit (GRS) mbH, Brunswick (Germany); Steininger, W. [Project Management Agency Karslruhe Water Technology and Waste Management, PTKA-WTE, Karlsruhe Institute of Technology, KIT, Eggenstein-Leopoldshafen (Germany); Miller, B. [AMEC, The Renaissance Center, Warrington (United Kingdom)

2015-06-15

A workshop was held in September 2012 in Braunschweig, Germany, to discuss the potential for natural and anthropogenic analogue studies to contribute to safety cases for radioactive waste repositories constructed in salt formations. Presentations were given on many analogue sites and systems from different countries. Discussions at the workshop then addressed the following aspects that are particularly relevant to the safety concept for radioactive waste disposal in salt: (1) the long-term integrity of rock salt formations, (2) the integrity of technical barriers, and (3) microbial, chemical and transport processes. A diverse range of natural systems were discussed as potential analogues for the integrity of rock salt. These included the deformation of anhydrite layers in rock salt; the response of rock salt to mechanical and thermal loads; and the isotopic signatures of syngenetic waters contained in fluid inclusions. Some anthropogenic examples drawn from the oil and gas industries, and from hazardous waste disposal, were proposed as analogues for the integrity of (geo)technical barriers. A broad range of studies on natural and anthropogenic salt-brine systems were identified as potential analogues for the radionuclide sorption and (co)precipitation process that may take place in the repository near and far fields, as well as for understanding the significance of hydrocarbons and microbial processes. It was evident from discussions at the workshop that there are some specific technical issues that may benefit from further analogue study, particularly the compaction of crushed salt backfill, the viability of microbes in the near-field, the stability of plugs and seals, the deformation of anhydrite, and isotope signatures in fluid inclusions. (authors)

People's perception of LILW repository

International Nuclear Information System (INIS)

Zeleznik, Nadja; Polic, Marko

2002-01-01

our surveys indicated significant differences in answers of experts and lay persons, mainly regarding evaluation of the consequences of repository construction. Although both groups believed that the land in the repository vicinity would lose its value, lay persons emphasized fear of the radioactivity as one of the most important factor, while experts emphasized increased safety due to the constant monitoring. For all of them assurance of safety, and constant informing of the public, were the main conditions of public acceptance of the repository. According to experts' opinion financial compensation could improve the acceptability, while for lay persons it was not so important. Authors believe that the new siting process should assure sufficient information, transparency of actions, and the use of professional but understandable language in communicating with the public, as well as permanent participation of local representatives in the siting process, with the possibility of withdrawing their approval at any stage. The fear of radioactivity should also be taken into consideration, even though it may not have a rational basis. Also it must be borne in mind, that external factors should also be considered, e.g. origin of the waste, high tension between centre and periphery, foreign examples and practices. (author)

Learning Object Repositories

Science.gov (United States)

Lehman, Rosemary

2007-01-01

This chapter looks at the development and nature of learning objects, meta-tagging standards and taxonomies, learning object repositories, learning object repository characteristics, and types of learning object repositories, with type examples. (Contains 1 table.)

Fulfillment of the long-term safety functions by the different barriers during the main time frames after repository closure

International Nuclear Information System (INIS)

Preter, P. de; Lalieux, Ph.

2002-01-01

In general terms the basis long-term safety functions of a disposal system (i.e. the engineered barrier system, including the waste forms and the host rock) are the functions that the system as a whole or its constituents must fulfill in order to assure an adequate level of long-term radiological safety. The long-term safety functions of a disposal system constitute a generic and methodological tool that can be used in a double sense. In the first place these functions provide an a priori instrument for designing the system: the implementer must ensure that these safety functions are fulfilled by a series of robust system barriers and components. These functions can also be used as an a posteriori means to describe and assess in general terms the functioning of the system. In this way they are an important qualitative element to help to support the safety case and to identify further R and D priorities. By providing a general description of system functioning they are also a communication tool to stakeholders who are less familiar with the details of a safety case. Instead of limiting the description to a multi-barrier system, the safety functions enable to better explain how the different barriers contribute to one or more safety functions and by which processes this is performed. By doing so the system description moves from multi-barrier to multi-function. The aim of this paper is to provide such a general description of the system functioning for the Belgian case of deep disposal of high-level waste (mainly spent fuel or vitrified waste from fuel reprocessing) in the Boom Clay, o poorly-indurated argillaceous formation. From the detailed safety and performance evaluations the main time frames after repository closure are identified. Each time frame relates to a period during which the successive safety functions play a key role. Also, in each time frame the radiological impact on the environment is distinctly different. (authors)

Safety case for the disposal of spent nuclear fuel at Olkiluoto - Synthesis 2012

Energy Technology Data Exchange (ETDEWEB)

NONE

2012-12-15

TURVA-2012 is Posiva's safety case in support of the Preliminary Safety Analysis Report (PSAR 2012) and application for a construction licence for a spent nuclear fuel repository. Consistent with the Government Decisions-in- Principle, this foresees a repository developed in bedrock at the Olkiluoto site according to the KBS-3 method, designed to accept spent nuclear fuel from the lifetime operations of the Olkiluoto and Loviisa reactors. Synthesis 2012 presents a synthesis of Posiva Oy's Safety Case 'TURVA-2012' portfolio. It summarises the design basis for the repository at the Olkiluoto site, the assessment methodology and key results of performance and safety assessments. It brings together all the lines of argument for safety, evaluation of compliance with the regulatory requirements, and statement of confidence in long-term safety and Posiva's safety analyses. The TURVA-2012 safety case demonstrates that the proposed repository design provides a safe solution for the disposal of spent nuclear fuel, and that the performance and safety assessments are fully consistent with all the legal and regulatory requirements related to long-term safety as set out in Government Decree 736/2008 and in guidance from the nuclear regulator - the STUK. Moreover, Posiva considers that the level of confidence in the demonstration of safety is appropriate and sufficient to submit the construction licence application to the authorities. The assessment of long-term safety includes uncertainties, but these do not affect the basic conclusions on the long-term safety of the repository. (orig.)

Perspectives on radioactive waste repository monitoring. Confirmation, compliance, confidence building, and societal vigilance

Energy Technology Data Exchange (ETDEWEB)

Bergmans, Anne [Antwerp Univ. (Belgium). Research Group Society and Environment; Elam, Mark [Gothenburg Univ. (Sweden); Simmons, Peter [East Anglia Univ., Norwich (United Kingdom); Sundqvist, Goeran [Oslo Univ. (Norway)

2012-12-15

Monitoring is now widely seen as a necessary part of programmes for the geological disposal of radioactive waste. However, we find different perspectives on the nature and role of monitoring. Among technical experts it is viewed firstly as a matter of performance confirmation, a tool for validating the safety case underlying repository construction. Among concerned citizens we find a view of monitoring as enabling the critical scrutiny of safety, an instrument for acknowledging uncertainties and detecting emergent problems in a repository. After outlining differing views on questions of whether, why, what, where and for how long to monitor we discuss monitoring in light of constant vigilance as a technical and moral principle of nuclear safety. We suggest that ''how much monitoring'' and ''how should it be organised'' are societal questions and as such need to be broadly discussed.

Evaluation of Nuclide Release Scenarios for a Hypothetical LILW Repository

International Nuclear Information System (INIS)

Lee, Youn Myoung; Jeong, Jong Tae

2010-11-01

A program for the safety assessment and performance evaluation of a low- and intermediate-level radioactive waste (LILW) repository system has been developed. Utilizing GoldSim (GoldSim, 2006), the program evaluates nuclide release and transport into the geosphere and biosphere under various disruptive natural and manmade events and scenarios that can occur after a waste package failure. We envisaged and illustrated these events and scenarios as occurring after the closure of a hypothetical LILW repository, and they included the degradation of various manmade barriers, pumping well drilling, and natural disruptions such as the sudden formation of a preferential flow pathway in the far-field area of the repository. Possible enhancement of nuclide transport facilitated by colloids or chelating agents is also dealt with. We used the newly-developed GoldSim template program, which is capable of various nuclide release scenarios and is greatly suited for simulating a potential repository given the geological circumstances in Korea, to create the detailed source term and near-field release scheme, various nuclide transport modes in the far-field geosphere area, and the biosphere transfer. Even though all parameter values applied to the hypothetical repository were assumed, the illustrative results, particularly the probabilistic calculations and sensitivity studies, may be informative under various scenarios

Preclosure Seismic Design Methodology for a Geologic Repository at Yucca Mountain

International Nuclear Information System (INIS)

K. Coppersmith

2004-01-01

This topical report describes the methodology and criteria that the U.S. Department of Energy (DOE) intends to use for preclosure seismic design of structures, systems, and components (SSCs) that are important to safety (ITS) in the geologic repository operations area. 10 Code of Federal Regulations (CFR) Part 63 [DIRS 156605], states that for a license to be issued for operation of a high-level radioactive waste repository, the U.S. Nuclear Regulatory Commission (NRC) must find that the facility will not constitute an unreasonable risk to the health and safety of the public (Section 63.41[c] [DIRS 156605]). Section 63.21(c)(5) [DIRS 156605] requires that a preclosure safety analysis (PCSA) be performed to ensure that the preclosure performance objectives (Section 63.111 [DIRS 156605]) have been met. The PCSA is a systematic examination of the site, the design, and the potential hazards (Section 63.102[f] [DIRS 156605]), including a comprehensive identification of potential event sequences. Potential naturally-occurring hazards include those event sequences that are initiated by earthquake ground motions or fault displacements due to earthquakes

Proposed format and content of license applications for deep geologic terminal repositories for radioactive material

International Nuclear Information System (INIS)

1978-01-01

Chapters are devoted to the following: introduction and general description; summary safety analysis; site characteristics; principal design criteria; repository design; operations systems; management of onsite generated waste; radiation protection; accident safety analysis; conduct of operations; operating controls and limits; and quality assurance

Analysis of the geological stability of a hypothetical radioactive waste repository in a bedded salt formation

International Nuclear Information System (INIS)

Tierney, M.S.; Lusso, F.; Shaw, H.R.

1978-01-01

This document reports on the development of mathematical models used in preliminary studies of the long-term safety of radioactive wastes deeply buried in bedded salt formations. Two analytical approaches to estimating the geological stability of a waste repository in bedded salt are described: (a) use of probabilistic models to estimate the a priori likelihoods of release of radionuclides from the repository through certain idealized natural and anthropogenic causes, and (b) a numerical simulation of certain feedback effects of emplacement of waste materials upon ground-water access to the repository's host rocks. These models are applied to an idealized waste repository for the sake of illustration

Template for safety reports with descriptive example

International Nuclear Information System (INIS)

1995-12-01

This report provides a template for future safety reports on long-term safety in support of important decisions and permit applications in connection with the construction of a deep repository system. The template aims at providing a uniform structure for describing long-term safety, after the repository has been closed and sealed. The availability of such a structure will simplify both preparation and review of the safety reports, and make it possible to follow how safety assessments are influenced by the progressively more detailed body of data that emerges. A separate section containing 'descriptive examples' has been appended to the template. This section illustrates what the different chapters of the template should contain. 279 refs

Template for safety reports with descriptive example

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-12-01

This report provides a template for future safety reports on long-term safety in support of important decisions and permit applications in connection with the construction of a deep repository system. The template aims at providing a uniform structure for describing long-term safety, after the repository has been closed and sealed. The availability of such a structure will simplify both preparation and review of the safety reports, and make it possible to follow how safety assessments are influenced by the progressively more detailed body of data that emerges. A separate section containing `descriptive examples` has been appended to the template. This section illustrates what the different chapters of the template should contain. 279 refs.

Radio ecological background for the isolation approach for the safety assessment of repositories

International Nuclear Information System (INIS)

Baltes, Bruno; Becker, Angela; Kindt, Anke

2007-01-01

A repository for radioactive waste should only be licensed if it poses no hazard to man and the environment. State of the art is the calculation of the potential radiation exposure of individuals in the surrounding area. A new concept has now been developed to assess the safe closure of radioactive waste in the isolating rock zone. Six criteria allow the quantification of the impact of the repository on the natural environmental conditions starting from the isolating rock zone over pore water and accessible water to the concentration in flora and fauna and to radiation exposure of humans placing the hitherto only criterion into a wider context. (orig.)

NAGRA - Long-term safety - The main task of deep repositories for radioactive wastes

International Nuclear Information System (INIS)

2015-10-01

This comprehensive brochure published by the Swiss National Cooperative for the Disposal of Radioactive Waste (NAGRA) examines the necessity for the safe disposal of radioactive wastes in Switzerland and discusses the requirements placed on such long-term waste depositories The effects of ionizing radiation on people and the protection provided by the deep repositories are examined. The construction of such deep repositories is looked at, as are the developments expected in the depositories over thousands of years. A comparison with natural occurrences is made and lessons to be learned from nature are discussed. Ideas for the marking of the depository sites are presented. A glossary of relevant terms completes the report

Hydrogeological modelling for assessment of radionuclide release scenarios for the repository system 2012

Energy Technology Data Exchange (ETDEWEB)

Hartley, L.; Hoek, J.; Swan, D.; Appleyard, P.; Baxter, S.; Roberts, D.; Simpson, T. [AMEC (United Kingdom)

2013-07-15

Posiva Oy is responsible for implementing the programme for geological disposal of spent nuclear fuel produced by its owners Teollisuuden Voima Oyj (TVO) and Fortum Power and Heat Oy in Finland. Olkiluoto in Eurajoki has been selected as the primary site for the repository, subject to further detailed investigation which is currently focused on the construction of an underground rock characterisation and research facility (the ONKALO). An essential part of the assessment of long-term safety of a repository is the analysis of groundwater flow since it is the only means of transport of radionuclides to the biosphere (besides human intrusion). The analysis of long-term safety for a KBS-3 concept requires as input a description of details of the groundwater flow around and through components of the engineered barrier system as well as details of the groundwater pathway to the biosphere during the current temperate climate period, as well as indications of behaviour under future climate periods such as glacial conditions. This report describes the groundwater flow modelling study performed to provide some of the necessary inputs required by Safety Assessment (i.e. radionuclide transport analysis). Underlying this study is the understanding of the site developed during the site investigations as summarised in the site descriptive model (SDM), and in particular the description of Olkiluoto Hydrogeological DFN model (Hydro-DFN). The main focus of this study is the temperate climate period, i.e. the evolution over the next 10,000 years, but the hydrogeological situation under various glacial climate conditions is also evaluated. Primary outputs of the study are repository performance measures relating to: the distributions of groundwater flow around the deposition holes; deposition tunnels and through the EDZ; flow-related transport resistance along groundwater pathways from the repository to the surface; and their the exit locations. Other analyses consider the

Hydrogeological modelling for assessment of radionuclide release scenarios for the repository system 2012

International Nuclear Information System (INIS)

Hartley, L.; Hoek, J.; Swan, D.; Appleyard, P.; Baxter, S.; Roberts, D.; Simpson, T.

2013-07-01

Posiva Oy is responsible for implementing the programme for geological disposal of spent nuclear fuel produced by its owners Teollisuuden Voima Oyj (TVO) and Fortum Power and Heat Oy in Finland. Olkiluoto in Eurajoki has been selected as the primary site for the repository, subject to further detailed investigation which is currently focused on the construction of an underground rock characterisation and research facility (the ONKALO). An essential part of the assessment of long-term safety of a repository is the analysis of groundwater flow since it is the only means of transport of radionuclides to the biosphere (besides human intrusion). The analysis of long-term safety for a KBS-3 concept requires as input a description of details of the groundwater flow around and through components of the engineered barrier system as well as details of the groundwater pathway to the biosphere during the current temperate climate period, as well as indications of behaviour under future climate periods such as glacial conditions. This report describes the groundwater flow modelling study performed to provide some of the necessary inputs required by Safety Assessment (i.e. radionuclide transport analysis). Underlying this study is the understanding of the site developed during the site investigations as summarised in the site descriptive model (SDM), and in particular the description of Olkiluoto Hydrogeological DFN model (Hydro-DFN). The main focus of this study is the temperate climate period, i.e. the evolution over the next 10,000 years, but the hydrogeological situation under various glacial climate conditions is also evaluated. Primary outputs of the study are repository performance measures relating to: the distributions of groundwater flow around the deposition holes; deposition tunnels and through the EDZ; flow-related transport resistance along groundwater pathways from the repository to the surface; and their the exit locations. Other analyses consider the

Safety case plan 2008

International Nuclear Information System (INIS)

2008-07-01

Following the guidelines set forth by the Ministry of Trade and Industry (now Ministry of Employment and Economy) Posiva is preparing to submit the construction license application for a spent fuel repository by the end of the year 2012. The long-term safety section supporting the license application is based on a safety case, which, according to the internationally adopted definition, is a compilation of the evidence, analyses and arguments that quantify and substantiate the safety and the level of expert confidence in the safety of the planned repository. In 2005, Posiva presented a plan to prepare such a safety case. The present report provides a revised plan of the safety case contents mentioned above. The update of the safety case plan takes into account the recommendations made by the Radiation and Nuclear Safety Authority (STUK) about improving the focus and further developing the plan. Accordingly, particular attention is given to the quality management of the safety case work, the management of uncertainties and the scenario methodology. The quality management is based on the ISO 9001:2000 standard process thinking enhanced with special features arising from STUK's YVL Guides. The safety case production process is divided into four main sub-processes. The conceptualisation and methodology sub-process defines the framework for the assessment. The critical data handling and modelling sub-process links Posiva's main technical and scientific activities to the production of the safety case. The assessment sub-process analyses the consequences of the evolution of the disposal system in various scenarios, classified either as part of the expected evolution or as disruptive scenarios. The compliance and confidence sub-process is responsible for final evaluation of compliance of the assessment results with the regulatory criteria and the overall confidence in the safety case. As in the previous safety case plan, the safety case will be based on several reports, but

Long-term Safety of a Geologic Repository: What does the public want to know?

International Nuclear Information System (INIS)

Kotra, Janet

2008-01-01

Janet Kotra of the U.S. NRC recounted the questions most frequently asked of the Commission's staff, concerning the long-term safety of the planned Yucca Mountain repository. The first question was related to who would decide whether the facility is safe enough. The next question linked to it is if regulatory staff is sufficiently qualified, and if they have access to independent sources of information or depend merely on the information provided by the implementer. The subsequent questions were about what criteria the regulator used and whether they were comprehensive enough. A separate group of questions addressed the safety of the technology, including the geologic features of the site, the engineered structures and the packaging of waste. Issues of documentation, operational control, long-term monitoring, oversight, and enforcement were also raised, along with the transparency of such processes and the accessibility of documents to the public. In addition, several questions were related to the fairness of decision making, public involvement, and the flexibility of the implementation process with regard to future knowledge and RD and D. Ms Kotra propounded that a single question was behind those cited above, namely whether one could rely on the regulator or not. Since it is not realistic to expect an overall consensus on the decision of the regulator agency, the goal should rather be to build confidence in the decision-making process. Key factors of confidence include openness on the part of the regulator and other decision-making institutions, and access to trustworthy representatives of these organisations

Safety upgrading of Novi Han Repository under IAEA TC Project BUL/4/005 achievements and future plans

International Nuclear Information System (INIS)

Stefanova, I.

2003-01-01

The report presents the safety upgrading of the Novi Han Repository under the IAEA TC Project BUL/4/005. The Project covers: identification of radionuclide inventory; characterisation of the disposal vaults; site characterisation; safety assessment; upgrading of the monitoring and radiation control; selection of treatment and conditioning processes and a conceptual design for a new waste processing and storage facility and other direct measures for safety improvement. The current inventory is identified and presented in the report. Schemes of the vault for solid wastes and vault for biological wastes are given, demonstrating reinforced concrete, stainless steel lining, and hydro insulation are presented. Several studies for safety assessment are made between 1997 and 2003. The operational safety assessment for disposal in existing facilities gives the annual risk for: spilling of waste package during upload (7.58.10 -9 ); spilling of waste package in transport accident (2.90.10 -9 ); fire scenario (3.50.10 - 1 3 ); radionuclide release due to flooding or earthquake (5.05.10 -4 ). The monitoring radiation control is upgraded according to the regulatory guidance and covered the site, restricted zone (1 km) and supervised zone (5 km). The types of analyses made are: Direct measurement of the dose rate -TLD; Direct measurements of the dose rate - portable surveillance monitors; In situ gamma spectrometry; Gamma spectrometry; Gross beta, gross alpha; Liquid scintillation spectrometry. The analyses show no transfer od radionuclides to the environment. The individual radiation control shows no evidence for specific radiation pathology. The operational radiation control service premises and transport vehicles. The following is measured: gamma dose rate; beta exposure; alpha exposure; neutron radiation; contamination level. Under the development is a detailed technical design supply of equipment for characterization of waste, including hot cell for control over high level

Performance assessment of Mochovce repository

Energy Technology Data Exchange (ETDEWEB)

Mrskova, A; Hanusik, V [Dept. of Accident Management and Risk Assessment, Vyskumny Ustav Jadrovych Elektrarni, Trnava (Slovakia)

2000-07-01

The near-surface disposal site at Mochovce is designed for low-level and intermediate level radioactive waste. It is a vault-type concrete structure housing the reinforced concrete containers as the final waste packages. This paper shortly presents the long-term safety analysis methods applied for the post-closure phase of the repository. The main aim of paper is description of the philosophy of analysis, development of the scenarios, their modeling and comparing of the results of normal evolution scenario, alternative scenario and intruders scenario for some radionuclides. (author)

Performance assessment of Mochovce repository

International Nuclear Information System (INIS)

Mrskova, A.; Hanusik, V.

2000-01-01

The near-surface disposal site at Mochovce is designed for low-level and intermediate level radioactive waste. It is a vault-type concrete structure housing the reinforced concrete containers as the final waste packages. This paper shortly presents the long-term safety analysis methods applied for the post-closure phase of the repository. The main aim of paper is description of the philosophy of analysis, development of the scenarios, their modeling and comparing of the results of normal evolution scenario, alternative scenario and intruders scenario for some radionuclides. (author)

A Safety Case Approach for Deep Geologic Disposal of DOE HLW and DOE SNF in Bedded Salt - 13350

Energy Technology Data Exchange (ETDEWEB)

Sevougian, S. David [Advanced Nuclear Energy Programs Group, Sandia National Laboratories, P.O. Box 5800, Albuquerque, NM 87185 (United States); MacKinnon, Robert J. [Advanced Nuclear Energy Programs Group, Sandia National Laboratories, P.O. Box 5800, Albuquerque, NM 87185 (United States); Leigh, Christi D. [Defense Waste Management Programs Group, Sandia National Laboratories, P.O. Box 5800, Albuquerque, NM 87185 (United States); Hansen, Frank D. [Geoscience Research and Applications Group, Sandia National Laboratories, P.O. Box 5800, Albuquerque, NM 87185 (United States)

2013-07-01

The primary objective of this study is to investigate the feasibility and utility of developing a defensible safety case for disposal of United States Department of Energy (U.S. DOE) high-level waste (HLW) and DOE spent nuclear fuel (SNF) in a conceptual deep geologic repository that is assumed to be located in a bedded salt formation of the Delaware Basin [1]. A safety case is a formal compilation of evidence, analyses, and arguments that substantiate and demonstrate the safety of a proposed or conceptual repository. We conclude that a strong initial safety case for potential licensing can be readily compiled by capitalizing on the extensive technical basis that exists from prior work on the Waste Isolation Pilot Plant (WIPP), other U.S. repository development programs, and the work published through international efforts in salt repository programs such as in Germany. The potential benefits of developing a safety case include leveraging previous investments in WIPP to reduce future new repository costs, enhancing the ability to effectively plan for a repository and its licensing, and possibly expediting a schedule for a repository. A safety case will provide the necessary structure for organizing and synthesizing existing salt repository science and identifying any issues and gaps pertaining to safe disposal of DOE HLW and DOE SNF in bedded salt. The safety case synthesis will help DOE to plan its future R and D activities for investigating salt disposal using a risk-informed approach that prioritizes test activities that include laboratory, field, and underground investigations. It should be emphasized that the DOE has not made any decisions regarding the disposition of DOE HLW and DOE SNF. Furthermore, the safety case discussed herein is not intended to either site a repository in the Delaware Basin or preclude siting in other media at other locations. Rather, this study simply presents an approach for accelerated development of a safety case for a potential

A Safety Case Approach for Deep Geologic Disposal of DOE HLW and DOE SNF in Bedded Salt - 13350

International Nuclear Information System (INIS)

Sevougian, S. David; MacKinnon, Robert J.; Leigh, Christi D.; Hansen, Frank D.

2013-01-01

The primary objective of this study is to investigate the feasibility and utility of developing a defensible safety case for disposal of United States Department of Energy (U.S. DOE) high-level waste (HLW) and DOE spent nuclear fuel (SNF) in a conceptual deep geologic repository that is assumed to be located in a bedded salt formation of the Delaware Basin [1]. A safety case is a formal compilation of evidence, analyses, and arguments that substantiate and demonstrate the safety of a proposed or conceptual repository. We conclude that a strong initial safety case for potential licensing can be readily compiled by capitalizing on the extensive technical basis that exists from prior work on the Waste Isolation Pilot Plant (WIPP), other U.S. repository development programs, and the work published through international efforts in salt repository programs such as in Germany. The potential benefits of developing a safety case include leveraging previous investments in WIPP to reduce future new repository costs, enhancing the ability to effectively plan for a repository and its licensing, and possibly expediting a schedule for a repository. A safety case will provide the necessary structure for organizing and synthesizing existing salt repository science and identifying any issues and gaps pertaining to safe disposal of DOE HLW and DOE SNF in bedded salt. The safety case synthesis will help DOE to plan its future R and D activities for investigating salt disposal using a risk-informed approach that prioritizes test activities that include laboratory, field, and underground investigations. It should be emphasized that the DOE has not made any decisions regarding the disposition of DOE HLW and DOE SNF. Furthermore, the safety case discussed herein is not intended to either site a repository in the Delaware Basin or preclude siting in other media at other locations. Rather, this study simply presents an approach for accelerated development of a safety case for a potential

Infiltration analysis for Abadia de Goias repository: numerical solution; Analise de infiltracao para o repositorio de Abadia de Goias: solucao numerica

Energy Technology Data Exchange (ETDEWEB)

Martin Alves, Antonio S. de [NUCLEN, Rio de Janeiro, RJ (Brazil); Passos, Aline M.M. dos [Universidade Federal Fluminense, Niteroi, RJ (Brazil)

1997-12-01

The safety analysis of a structure known as repository for medium activity wastes leads to investigating the physical phenomena connected to the water infiltration. This work shows succinctly an engineering approach to obtain numerical results for the model differential equations. One of these equations, related to the two-phase flow within the structure, is a nonlinear Riccati type, whose solution is only known for certain cases. For safety analysis and design purposes, the solution for the case of variable parameters is also advantageous when one aims some accident scenarios analysis. The utilization of numerical techniques allowed excellent results applied for the design of the Abadia de Goias repository. The case treated in this paper was one of those applied to the safety assessment of this repository. (author). 7 refs., 4 figs., 7 tabs.

Data report for the safety assessment SR-Site

International Nuclear Information System (INIS)

2010-12-01

This report compiles, documents, and qualifies input data identified as essential for the long-term safety assessment of a KBS-3 repository, and forms an important part of the reporting of the safety assessment project SR-Site. The input data concern the repository system, broadly defined as the deposited spent nuclear fuel, the engineered barriers surrounding it, the host rock, and the biosphere in the proximity of the repository. The input data also concern external influences acting on the system, in terms of climate related data. Data are provided for a selection of relevant conditions and are qualified through traceable standardised procedures

Data report for the safety assessment SR-Site

Energy Technology Data Exchange (ETDEWEB)

2010-12-15

This report compiles, documents, and qualifies input data identified as essential for the long-term safety assessment of a KBS-3 repository, and forms an important part of the reporting of the safety assessment project SR-Site. The input data concern the repository system, broadly defined as the deposited spent nuclear fuel, the engineered barriers surrounding it, the host rock, and the biosphere in the proximity of the repository. The input data also concern external influences acting on the system, in terms of climate related data. Data are provided for a selection of relevant conditions and are qualified through traceable standardised procedures

Siting regions for deep geological repositories. Why just here?

International Nuclear Information System (INIS)

Rieser, A.

2009-09-01

This report helps to the popularization of the Nagra works accomplished for the management and disposal of the radioactive wastes in Switzerland. The programme for management and disposal of the radioactive wastes are extensively determined by regulations. Protection of mankind and environment is the primary objective. The basic storage process is considered as having been solved. The question addressed in the report is where the facility has to be built; the site selection procedure includes five steps: 1) according to their type the wastes have to be allocated to two different repositories: for low- and intermediate-level wastes (L/ILW), and for high-level and alpha-toxic wastes (HLW); 2) the safety concept for both repositories and the requirements on the geology have to be determined; 3) large suitable geological-tectonic zones must be found where repositories could be built; 4) in these geological zones a suitable host rock has to be identified; 5) the most important spatial geological conditions of the host rock (minimum depth with respect to surface erosion, maximum depth in terms of engineering requirements, lateral extent) have to be identified. Based on these criteria, three suitable siting regions for a HLW repository were found in the North of Switzerland. The preferred host rock is Opalinus clay because of its very low permeability; it is therefore an excellent barrier against nuclide transport. In the three proposed siting regions, Opalinus clay is present in sufficient volumes at a suitable depth. For a L/ILW repository six different possible siting regions were identified, five in Northern Switzerland and one in Central Switzerland. In the three siting regions found for a possible HLW repository, it would also be possible to built a combined repository for both HLW and L/ILW wastes

Summary of four release consequence analyses for hypothetical nuclear waste repositories in salt and granite