WorldWideScience

Sample records for repository performance assessment

  1. Modelling saline intrusion for repository performance assessment

    International Nuclear Information System (INIS)

    Jackson, C.P.

    1989-04-01

    UK Nirex Ltd are currently considering the possibility of disposal of radioactive waste by burial in deep underground repositories. The natural pathway for radionuclides from such a repository to return to Man's immediate environment (the biosphere) is via groundwater. Thus analyses of the groundwater flow in the neighbourhood of a possible repository, and consequent radionuclide transport form an important part of a performance assessment for a repository. Some of the areas in the UK that might be considered as possible locations for a repository are near the coast. If a repository is located in a coastal region seawater may intrude into the groundwater flow system. As seawater is denser than fresh water buoyancy forces acting on the intruding saline water may have significant effects on the groundwater flow system, and consequently on the time for radionuclides to return to the biosphere. Further, the chemistry of the repository near-field may be strongly influenced by the salinity of the groundwater. It is therefore important for Nirex to have a capability for reliably modelling saline intrusion to an appropriate degree of accuracy in order to make performance assessments for a repository in a coastal region. This report describes work undertaken in the Nirex Research programme to provide such a capability. (author)

  2. 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)

  3. 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)

  4. Performance Assessment Strategy Plan for the Geologic Repository Program

    International Nuclear Information System (INIS)

    1990-01-01

    Performance assessment is a major constituent of the program being conducted by the US Department of Energy (DOE) to develop a geologic repository. Performance assessment is the set of activities needed for quantitative evaluations to assess compliance with the performance requirements in the regulations for a geologic repository and to support the development of the repository. The strategy for these evaluations has been documented in the Performance Assessment Strategy Plan (DOE, 1989). The implementation of the performance assessment strategy is defined in this document. This paper discusses the scope and objectives of the implementation plan, the relationship of the plan to other program plans, summarizes the performance assessment areas and the integrated strategy of the performance assessment program. 1 fig., 3 tabs

  5. 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)

  6. Performance assessment plans and methods for the Salt Repository Project

    International Nuclear Information System (INIS)

    1984-08-01

    This document presents the preliminary plans and anticipated methods of the Salt Repository Project (SRP) for assessing the postclosure and radiological aspects of preclosure performance of a nuclear waste repository in salt. This plan is intended to be revised on an annual basis. The emphasis in this preliminary effort is on the method of conceptually dividing the system into three subsystems (the very near field, the near field, and the far field) and applying models to analyze the behavior of each subsystem and its individual components. The next revision will contain more detailed plans being developed as part of Site Characterization Plan (SCP) activities. After a brief system description, this plan presents the performance targets which have been established for nuclear waste repositories by regulatory agencies (Chapter 3). The SRP approach to modeling, including sensitivity and uncertainty techniques is then presented (Chapter 4). This is followed by a discussion of scenario analysis (Chapter 5), a presentation of preliminary data needs as anticipated by the SRP (Chapter 6), and a presentation of the SRP approach to postclosure assessment of the very near field, the near field, and the far field (Chapters 7, 8, and 9, respectively). Preclosure radiological assessment is discussed in Chapter 10. Chapter 11 presents the SRP approach to code verification and validation. Finally, the Appendix lists all computer codes anticipated for use in performance assessments. The list of codes will be updated as plans are revised

  7. SKI SITE-94, deep repository performance assessment project, summary

    International Nuclear Information System (INIS)

    1999-01-01

    SITE-94 is a comprehensive performance assessment exercise for a hypothetical repository for spent nuclear fuel at a real site in Sweden. SITE-94 was carried out to develop the capability and tools to enable Swedish Nuclear Power Inspectorate (SKI) to review fully the proposals for a deep repository which are expected to be made by the Swedish Nuclear Fuel and Waste Management Company, SKB (the implementor). Sweden is one of the leading countries in the research and development of geological disposal of radioactive waste. The developed methodology for performance assessment has attracted interests from other countries. The Summary of the main report of the SITE-94 project is translated here into Japanese to allow to make the information on the methodology and the related issues available among Japanese concerned. (author)

  8. SKI SITE-94. Deep Repository Performance Assessment Project. Summary

    International Nuclear Information System (INIS)

    1997-02-01

    The function of SITE-94 is to provide the Swedish Nuclear Power Inspectorate (SKI) with the capacity and supporting knowledge needed for reviewing the Swedish nuclear industry's R and D programs and for reviewing license applications, as stipulated in Swedish legislation. The report is structured as a Performance Assessment exercise needed for input to decisions regarding repository safety, but the SITE-94 is neither a safety assessment nor a model for future assessments to be undertaken by the prospective licensee. The specific project objectives of SITE-94 comprise site evaluation, performance assessment methodology, canister integrity and radionuclide release and transport calculations. The main report (SKI-R--96-36) gives a detailed description of the many inter-related studies undertaken as part of the research project, while the present report presents a condensed summary of the main report. 46 refs

  9. Some considerations for validation of repository performance assessment models

    International Nuclear Information System (INIS)

    Eisenberg, N.

    1991-01-01

    Validation is an important aspect of the regulatory uses of performance assessment. A substantial body of literature exists indicating the manner in which validation of models is usually pursued. Because performance models for a nuclear waste repository cannot be tested over the long time periods for which the model must make predictions, the usual avenue for model validation is precluded. Further impediments to model validation include a lack of fundamental scientific theory to describe important aspects of repository performance and an inability to easily deduce the complex, intricate structures characteristic of a natural system. A successful strategy for validation must attempt to resolve these difficulties in a direct fashion. Although some procedural aspects will be important, the main reliance of validation should be on scientific substance and logical rigor. The level of validation needed will be mandated, in part, by the uses to which these models are put, rather than by the ideal of validation of a scientific theory. Because of the importance of the validation of performance assessment models, the NRC staff has engaged in a program of research and international cooperation to seek progress in this important area. 2 figs., 16 refs

  10. Performance Assessments of Generic Nuclear Waste Repositories in Shale

    Science.gov (United States)

    Stein, E. R.; Sevougian, S. D.; Mariner, P. E.; Hammond, G. E.; Frederick, J.

    2017-12-01

    Simulations of deep geologic disposal of nuclear waste in a generic shale formation showcase Geologic Disposal Safety Assessment (GDSA) Framework, a toolkit for repository performance assessment (PA) whose capabilities include domain discretization (Cubit), multiphysics simulations (PFLOTRAN), uncertainty and sensitivity analysis (Dakota), and visualization (Paraview). GDSA Framework is used to conduct PAs of two generic repositories in shale. The first considers the disposal of 22,000 metric tons heavy metal of commercial spent nuclear fuel. The second considers disposal of defense-related spent nuclear fuel and high level waste. Each PA accounts for the thermal load and radionuclide inventory of applicable waste types, components of the engineered barrier system, and components of the natural barrier system including the host rock shale and underlying and overlying stratigraphic units. Model domains are half-symmetry, gridded with Cubit, and contain between 7 and 22 million grid cells. Grid refinement captures the detail of individual waste packages, emplacement drifts, access drifts, and shafts. Simulations are run in a high performance computing environment on as many as 2048 processes. Equations describing coupled heat and fluid flow and reactive transport are solved with PFLOTRAN, an open-source, massively parallel multiphase flow and reactive transport code. Additional simulated processes include waste package degradation, waste form dissolution, radioactive decay and ingrowth, sorption, solubility, advection, dispersion, and diffusion. Simulations are run to 106 y, and radionuclide concentrations are observed within aquifers at a point approximately 5 km downgradient of the repository. Dakota is used to sample likely ranges of input parameters including waste form and waste package degradation rates and properties of engineered and natural materials to quantify uncertainty in predicted concentrations and sensitivity to input parameters. Sandia National

  11. Sorption data base for performance assessment of radwaste repository

    International Nuclear Information System (INIS)

    Cho, Y.H.; Jung, J.; Lee, J.K.; Hahn, P.S.

    2001-01-01

    Sorption data base (SDB) provides readily available data for the performance assessment of radwaste repository when site-specific data are not available and/or more reference data are needed. The software developed in the Korea Atomic Energy Research Institute (KAERI), SDB-21C, is a graphic user interface (GUI) program that provides efficient and user friendly tools for evaluating the large amount of sorption data. The data base of distribution coefficients compiled in the program contains about 11,000 Nuclear Energy Agency (NEA) data and 2,000 KAERI data up to now while the addition of new data is under progress. Furthermore, the parametric model and its compiled data sets are also included in SDB-21C. (author)

  12. Validation of geotechnical software for repository performance assessment

    International Nuclear Information System (INIS)

    LeGore, T.; Hoover, J.D.; Khaleel, R.; Thornton, E.C.; Anantatmula, R.P.; Lanigan, D.C.

    1989-01-01

    An important step in the characterization of a high level nuclear waste repository is to demonstrate that geotechnical software, used in performance assessment, correctly models validation. There is another type of validation, called software validation. It is based on meeting the requirements of specifications documents (e.g. IEEE specifications) and does not directly address the correctness of the specifications. The process of comparing physical experimental results with the predicted results should incorporate an objective measure of the level of confidence regarding correctness. This paper reports on a methodology developed that allows the experimental uncertainties to be explicitly included in the comparison process. The methodology also allows objective confidence levels to be associated with the software. In the event of a poor comparison, the method also lays the foundation for improving the software

  13. Approach to geologic repository post closure system performance assessment

    International Nuclear Information System (INIS)

    Pahwa, S.B.; Felton, W.; Duguid, J.O.

    1992-01-01

    An essential part of the license application for a geologic repository will be the demonstration of compliance with the standards set by the Environmental Protection Agency. The performance assessments that produce the demonstration must rely on models of various levels of detail. The most detailed of these models are needed for understanding thoroughly the complex physical and chemical processes affecting the behavior of the system. For studying the behavior of major components of the system, less detailed models are often useful. For predicting the behavior of the total system, models of a third kind may be needed. These models must cover all the important processes that contribute to the behavior of the system, because they must estimate the behavior under all significant conditions for 10,000 years. In addition, however, computer codes that embody these models must calculate very rapidly because of the EPA standard's requirement for probabilistic estimates, which will be produced by sampling thousands of times from probability distributions of parameters. For this reason, the total-system models must be less complex than the detailed-process and subsystem models. The total-system performance is evaluated through modeling of the following components: Radionuclide release from the engineered-barrier system. Fluid flow in the geologic units. Radionuclide transport to the accessible environment. Radionuclide release to the accessible environment and dose to man

  14. Repository performance confirmation

    International Nuclear Information System (INIS)

    Hansen, Francis D.

    2011-01-01

    Yucca Mountain license application identified a broad suite of monitoring activities. A revision of the plan was expected to winnow the number of activities down to a manageable size. As a result, an objective process for the next stage of performance confirmation planning was developed as an integral part of an overarching long-term testing and monitoring strategy. The Waste Isolation Pilot Plant compliance monitoring program at once reflects its importance to stakeholders while demonstrating adequate understanding of relevant monitoring parameters. The compliance criteria were stated by regulation and are currently monitored as part of the regulatory rule for disposal. At the outset, the screening practice and parameter selection were not predicated on a direct or indirect correlation to system performance metrics, as was the case for Yucca Mountain. Later on, correlation to performance was established, and the Waste Isolation Pilot Plant continues to monitor ten parameters originally identified in the compliance certification documentation. The monitoring program has proven to be effective for the technical intentions and societal or public assurance. The experience with performance confirmation in the license application process for Yucca Mountain helped identify an objective, quantitative methodology for this purpose. Revision of the existing plan would be based on findings of the total system performance assessment. Identification and prioritization of confirmation activities would then derive from performance metrics associated with performance assessment. Given the understanding of repository performance confirmation, as reviewed in this paper, it is evident that the performance confirmation program for the Yucca Mountain project could be readily re-engaged if licensing activities resumed.

  15. Site specificity of biosphere parameter values in performance assessments of near-surface repositories

    International Nuclear Information System (INIS)

    Zeevaert, Th.; Volckaert, G.; Vandecasleele

    1993-01-01

    The contribution is dealing with the performance assessment model for near surface repositories in Belgium. It consists of four submodels called: site, aquifer, biosphere and dose. For some characteristic radionuclides, results of the study are shown for a typical site, and differences in doses assessed with the generic approach discussed. Shortcomings are indicated

  16. Hydrological performance assessment on siting the high level radioactive waste repository

    International Nuclear Information System (INIS)

    Guo Yonghai; Liu Shufen; Wang Ju; Wang Zhiming; Su Rui; Lv Chuanhe; Zong Zihua

    2007-01-01

    Based on the research experiences in China and some developed countries in the world, the processes and methods on hydrological performance assessment for the siting of high radioactive repository are discussed in this paper. The methods and contents of hydrological performance assessment are discussed respectively for region, area and site hydrological investigation stages. At the same time, the hydrological performance assessment of the potential site for high level radioactive waste in China is introduced. (authors)

  17. Performance assessment for Nuclear Power Plant Krsko intermediate and low-level rad-waste repository

    International Nuclear Information System (INIS)

    Jelavic, V.; Skanata, D.; Plecjas, I.

    1990-01-01

    Performance safety Assessment for NPP Krsko radwaste repository was performed (LLW/ILW). Shallow land and tunnel type concept were analyzed. Because it was based on two unknown referent sites, one for the shallow land concept and the other for the tunnel type, analysis was generic in nature. Scenario selecting process and consequence analysis were performed by using deterministic approach. Results for both concepts of disposal suggests that proposed NPP Krsko radwaste repository reference site and disposal technology will fully meet radiation limits imposed by the Yugoslav regulations and ICRP guidelines. (author)

  18. Workshop on rock mechanics issues in repository design and performance assessment

    International Nuclear Information System (INIS)

    1996-04-01

    The Center for Nuclear Waste Regulatory Analyses organized and hosted a workshop on ''Rock Mechanics Issues in Repository Design and Performance Assessment'' on behalf its sponsor the U.S. Nuclear Regulatory Commission (NRC). This workshop was held on September 19- 20, 1994 at the Holiday Inn Crowne Plaza, Rockville, Maryland. The objectives of the workshop were to stimulate exchange of technical information among parties actively investigating rock mechanics issues relevant to the proposed high-level waste repository at Yucca Mountain and identify/confirm rock mechanics issues important to repository design and performance assessment The workshop contained three technical sessions and two panel discussions. The participants included technical and research staffs representing the NRC and the Department of Energy and their contractors, as well as researchers from the academic, commercial, and international technical communities. These proceedings include most of the technical papers presented in the technical sessions and the transcripts for the two panel discussions

  19. 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

  20. A performance assessment review tool for the proposed radioactive waste repository at Yucca Mountain, Nevada, USA

    International Nuclear Information System (INIS)

    Mohanty, Sitakanta; Codell, Richard

    2000-01-01

    The U.S. Nuclear Regulatory Commission (NRC), with the assistance of the Center for Nuclear Waste Regulatory Analyses, has developed a Total-system Performance Assessment (TPA) Code to assist in evaluating the performance of the Yucca Mountain (YM) High-Level Waste Repository in Nevada, proposed by the U.S. Department of Energy (DOE). The proposed YM repository would be built in a thick sequence of partially saturated volcanic tuff above the water table. Among the unique challenges of this environment are (1) the transport of radionuclides would take place partially through highly heterogeneous unsaturated rock; (2) the waste packages (WPs) would be generally exposed to oxidizing conditions, and (3) water either infiltrating from the surface or recirculating because of decay heat may drip onto the WPs. Tools such as the TPA code and embedded techniques for evaluating YM performance are aimed at (1) determining the parameters and key parts of the repository system that have the most influence on repository performance; (2) performing alternative conceptual models studies, especially with bounding models; (3) estimating the relative importance of the physical phenomena that lead to human exposure to radionuclides; and (4) improving NRC staff capabilities in performance assessment and associated license application reviews. This paper presents an overview of the NRC conceptual framework, approach to conducting system-level sensitivity analyses for determining influential parameters, and alternative conceptual model studies to investigate the effect of model uncertainties. (author)

  1. Demonstration of a repository performance assessment capability at the US Nuclear Regulatory Commission

    International Nuclear Information System (INIS)

    Codell, R.; Eisenberg, N.; McCartin, T.; Park, J.

    1991-01-01

    In order to better review licensing submittals for a High-Level Waste Repository, the US Nuclear Regulatory Commission staff has expanded and improved its capability to conduct performance assessments. A demonstration of this capability used the limited data from Yucca Mountain, Nevada to investigate a small set of scenario classes. Models of release and transport of radionuclides from a repository via the groundwater and direct release pathways provided preliminary estimates of releases to the accessible environment for a 10,000 year simulation time. Latin hypercube sampling of input parameters was used to express results as distributions and to investigate model sensitivities. This methodology demonstration should not be interpreted as an estimate of performance of the proposed repository at Yucca Mountain, Nevada

  2. The implications of heterogeneity for repository performance assessments

    International Nuclear Information System (INIS)

    Jackson, C.P.; Porter, J.D.; Morris, S.T.; Herbert, A.W.

    1991-01-01

    We outline the current views of the Nirex Disposal Safety Assessment Team on heterogeneity, we describe the pragmatic approach to modelling the consequences of heterogeneity that is being currently used, we present work that is being undertaken in the Nirex Safety Assessment Research Programme to develop improved models and we discuss the implications of heterogeneity for site investigation. We point out the need to develop simple models for use in probabilistic analyses. Heterogeneity leads to dispersion, which is currently modelled using a simple diffusion-like model. We discuss the differences between structured heterogeneity, such as fracture zones, and random heterogeneity. We consider that the geostatistical approach to modelling random heterogeneity is probably that most suitable for the needs of Nirex. More measurements are needed in order to characterize heterogeneous media than to characterize homogeneous media. 18 refs., 4 figs

  3. Improved metamodel-based importance sampling for the performance assessment of radioactive waste repositories

    International Nuclear Information System (INIS)

    Cadini, F.; Gioletta, A.; Zio, E.

    2015-01-01

    In the context of a probabilistic performance assessment of a radioactive waste repository, the estimation of the probability of exceeding the dose threshold set by a regulatory body is a fundamental task. This may become difficult when the probabilities involved are very small, since the classically used sampling-based Monte Carlo methods may become computationally impractical. This issue is further complicated by the fact that the computer codes typically adopted in this context requires large computational efforts, both in terms of time and memory. This work proposes an original use of a Monte Carlo-based algorithm for (small) failure probability estimation in the context of the performance assessment of a near surface radioactive waste repository. The algorithm, developed within the context of structural reliability, makes use of an estimated optimal importance density and a surrogate, kriging-based metamodel approximating the system response. On the basis of an accurate analytic analysis of the algorithm, a modification is proposed which allows further reducing the computational efforts by a more effective training of the metamodel. - Highlights: • We tackle uncertainty propagation in a radwaste repository performance assessment. • We improve a kriging-based importance sampling for estimating failure probabilities. • We justify the modification by an analytic, comparative analysis of the algorithms. • The probability of exceeding dose thresholds in radwaste repositories is estimated. • The algorithm is further improved reducing the number of its free parameters

  4. Elicitation and use of expert judgment in performance assessment for high-level radioactive waste repositories

    International Nuclear Information System (INIS)

    Bonano, E.J.; Hora, S.C.; Keeney, R.L.; von Winterfeldt, D.

    1990-05-01

    This report presents the concept of formalizing the elicitation and use of expert judgment in the performance assessment of high-level radioactive waste (HLW) repositories in deep geologic formations. The report begins with a discussion of characteristics (advantages and disadvantages) of formalizing expert judgment examples of previous uses of expert judgment in radioactive waste programs, criteria that can assist in deciding when to formalize expert judgment, and the relationship of formal use of expert judgment to data collection and modeling. The current state of the art with respect to the elicitation, use, and communication of formal expert judgment is presented. The report concludes with a discussion on potential applications of formal expert judgment in performance assessment of HLW repositories. 93 refs

  5. A suggested approach toward measuring sorption and applying sorption data to repository performance assessment

    International Nuclear Information System (INIS)

    Rundberg, R.S.

    1992-01-01

    The prediction of radionuclide migration for the purpose of assessing the safety of a nuclear waste repository will be based on a collective knowledge of hydrologic and geochemical properties of the surrounding rock and groundwater. This knowledge along with assumption about the interactions of radionuclides with groundwater and minerals form the scientific basis for a model capable of accurately predicting the repository's performance. Because the interaction of radionuclides in geochemical systems is known to be complicated, several fundamental and empirical approaches to measuring the interaction between radionuclides and the geologic barrier have been developed. The approaches applied to the measurement of sorption involve the use of pure minerals, intact, or crushed rock in dynamic and static experiments. Each approach has its advantages and disadvantages. There is no single best method for providing sorption data for performance assessment models which can be applied without invoking information derived from multiple experiments. 53 refs., 12 figs

  6. 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.

  7. 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.

  8. Role of natural analogs in performance assessment of nuclear waste repositories

    International Nuclear Information System (INIS)

    Sagar, B.; Wittmeyer, G.W.

    1995-01-01

    Mathematical models of the flow of water and transport of radionuclides in porous media will be used to assess the ability of deep geologic repositories to safely contain nuclear waste. These models must, in some sense, be validated to ensure that they adequately describe the physical processes occurring within the repository and its geologic setting. Inasmuch as the spatial and temporal scales over which these models must be applied in performance assessment are very large, validation of these models against laboratory and small-scale field experiments may be considered inadequate. Natural analogs may provide validation data that are representative of physico-chemical processes that occur over spatial and temporal scales as large or larger than those relevant to repository design. The authors discuss the manner in which natural analog data may be used to increase confidence in performance assessment models and conclude that, while these data may be suitable for testing the basic laws governing flow and transport, there is insufficient control of boundary and initial conditions and forcing functions to permit quantitative validation of complex, spatially distributed flow and transport models. The authors also express their opinion that, for collecting adequate data from natural analogs, resources will have to be devoted to them that are much larger than are devoted to them at present

  9. Workshop on rock mechanics issues in repository design and performance assessment

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-04-01

    The Center for Nuclear Waste Regulatory Analyses organized and hosted a workshop on ``Rock Mechanics Issues in Repository Design and Performance Assessment`` on behalf its sponsor the U.S. Nuclear Regulatory Commission (NRC). This workshop was held on September 19- 20, 1994 at the Holiday Inn Crowne Plaza, Rockville, Maryland. The objectives of the workshop were to stimulate exchange of technical information among parties actively investigating rock mechanics issues relevant to the proposed high-level waste repository at Yucca Mountain and identify/confirm rock mechanics issues important to repository design and performance assessment The workshop contained three technical sessions and two panel discussions. The participants included technical and research staffs representing the NRC and the Department of Energy and their contractors, as well as researchers from the academic, commercial, and international technical communities. These proceedings include most of the technical papers presented in the technical sessions and the transcripts for the two panel discussions. Selected papers have been indexed separately for inclusion the Energy Science and Technology Database.

  10. The use of performance assessments in Yucca Mountain repository waste package design activities

    International Nuclear Information System (INIS)

    Jardine, L.J.

    1990-01-01

    The Yucca Mountain Project is developing performance assessment approaches as part of the evaluations of the suitability of Yucca Mountain as a repository site. Lawrence Livermore National Laboratory is developing design concepts and the scientific performance assessment methodologies and techniques used for the waste package and engineered barrier system components. This paper presents an overview of the approach under development for postclosure performance assessments that will guide the conceptual design activities and assist in the site suitability evaluations. This approach includes establishing and modeling for the long time periods required by regulations: near-field environment characteristics surrounding the emplaced wastes; container materials performance responses; and waste form properties. All technical work is being done under a fully qualified quality assurance program

  11. Use of Groundwater Lifetime Expectancy for the Performance Assessment of Deep Geologic Radioactive Waste Repositories.

    Science.gov (United States)

    Cornaton, F.; Park, Y.; Normani, S.; Sudicky, E.; Sykes, J.

    2005-12-01

    Long-term solutions for the disposal of toxic wastes usually involve isolation of the wastes in a deep subsurface geologic environment. In the case of spent nuclear fuel, the safety of the host repository depends on two main barriers: the engineered barrier and the natural geological barrier. If radionuclide leakage occurs from the engineered barrier, the geological medium represents the ultimate barrier that is relied upon to ensure safety. Consequently, an evaluation of radionuclide travel times from the repository to the biosphere is critically important in a performance assessment analysis. In this study, we develop a travel time framework based on the concept of groundwater lifetime expectancy as a safety indicator. Lifetime expectancy characterizes the time radionuclides will spend in the subsurface after their release from the repository and prior to discharging into the biosphere. The probability density function of lifetime expectancy is computed throughout the host rock by solving the backward-in-time solute transport equation subject to a properly posed set of boundary conditions. It can then be used to define optimal repository locations. In a second step, the risk associated with selected sites can be evaluated by simulating an appropriate contaminant release history. The proposed methodology is applied in the context of a typical Canadian Shield environment. Based on a statistically-generated three-dimension network of fracture zones embedded in the granitic host rock, the sensitivity and the uncertainty of lifetime expectancy to the hydraulic and dispersive properties of the fracture network, including the impact of conditioning via their surface expressions, is computed in order to demonstrate the utility of the methodology.

  12. Repository Integration Program: RIP performance assessment and strategy evaluation model theory manual and user's guide

    International Nuclear Information System (INIS)

    1995-11-01

    This report describes the theory and capabilities of RIP (Repository Integration Program). RIP is a powerful and flexible computational tool for carrying out probabilistic integrated total system performance assessments for geologic repositories. The primary purpose of RIP is to provide a management tool for guiding system design and site characterization. In addition, the performance assessment model (and the process of eliciting model input) can act as a mechanism for integrating the large amount of available information into a meaningful whole (in a sense, allowing one to keep the ''big picture'' and the ultimate aims of the project clearly in focus). Such an integration is useful both for project managers and project scientists. RIP is based on a '' top down'' approach to performance assessment that concentrates on the integration of the entire system, and utilizes relatively high-level descriptive models and parameters. The key point in the application of such a ''top down'' approach is that the simplified models and associated high-level parameters must incorporate an accurate representation of their uncertainty. RIP is designed in a very flexible manner such that details can be readily added to various components of the model without modifying the computer code. Uncertainty is also handled in a very flexible manner, and both parameter and model (process) uncertainty can be explicitly considered. Uncertainty is propagated through the integrated PA model using an enhanced Monte Carlo method. RIP must rely heavily on subjective assessment (expert opinion) for much of its input. The process of eliciting the high-level input parameters required for RIP is critical to its successful application. As a result, in order for any project to successfully apply a tool such as RIP, an enormous amount of communication and cooperation must exist between the data collectors, the process modelers, and the performance. assessment modelers

  13. Performance assessments of nuclear waste repositories--A dialogue on their value and limitations

    Science.gov (United States)

    Ewing, Rodney C.; Tierney, Martin S.; Konikow, Leonard F.; Rechard, Rob P.

    1999-01-01

    Performance Assessment (PA) is the use of mathematical models to simulate the long-term behavior of engineered and geologic barriers in a nuclear waste repository; methods of uncertainty analysis are used to assess effects of parametric and conceptual uncertainties associated with the model system upon the uncertainty in outcomes of the simulation. PA is required by the U.S. Environmental Protection Agency as part of its certification process for geologic repositories for nuclear waste. This paper is a dialogue to explore the value and limitations of PA. Two “skeptics” acknowledge the utility of PA in organizing the scientific investigations that are necessary for confident siting and licensing of a repository; however, they maintain that the PA process, at least as it is currently implemented, is an essentially unscientific process with shortcomings that may provide results of limited use in evaluating actual effects on public health and safety. Conceptual uncertainties in a PA analysis can be so great that results can be confidently applied only over short time ranges, the antithesis of the purpose behind long-term, geologic disposal. Two “proponents” of PA agree that performance assessment is unscientific, but only in the sense that PA is an engineering analysis that uses existing scientific knowledge to support public policy decisions, rather than an investigation intended to increase fundamental knowledge of nature; PA has different goals and constraints than a typical scientific study. The “proponents” describe an ideal, sixstep process for conducting generalized PA, here called probabilistic systems analysis (PSA); they note that virtually all scientific content of a PA is introduced during the model-building steps of a PSA, they contend that a PA based on simple but scientifically acceptable mathematical models can provide useful and objective input to regulatory decision makers. The value of the results of any PA must lie between these two

  14. Sources/treatment of uncertainties in the performance assessment of geologic radioactive waste repositories

    International Nuclear Information System (INIS)

    Cranwell, R.M.

    1987-01-01

    Uncertainties in the performance assessment of geologic radioactive waste repositories have several sources. The more important ones include: 1) uncertainty in the conditions of a disposal system over the temporal scales set forth in regulations, 2) uncertainty in the conceptualization of the geohydrologic system, 3) uncertainty in the theoretical description of a given conceptual model of the system, 4) uncertainty in the development of computer codes to implement the solution of a mathematical model, and 5) uncertainty in the parameters and data required in the models and codes used to assess the long-term performance of the disposal system. This paper discusses each of these uncertainties and outlines methods for addressing these uncertainties

  15. SKI SITE-94. Deep Repository Performance Assessment Project Volume I and II

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-12-01

    The function of SITE-94 is to provide the Swedish Nuclear Power Inspectorate (SKI) with the capacity and supporting knowledge needed for reviewing the Swedish nuclear industry`s R and D programs and for reviewing license applications, as stipulated in Swedish legislation. The report is structured as a Performance Assessment exercise needed for input to decisions regarding repository safety, but the SITE-94 is not a safety assessment or a model for future assessments to be undertaken by the prospective licensee. The specific project objectives of SITE-94 comprise site evaluation, performance assessment methodology, canister integrity and radionuclide release and transport calculations. The report gives a detailed description of the many inter-related studies undertaken as part of the research project. As a general conclusion it is noted that the SITE-94 project has significantly advanced SKI`s capability of reviewing performance assessments and has led to the development of a tool-kit for carrying out assessment calculations, both in terms of numerical and scoping analyses. The study has highlighted a number of key issues for safety assessment (and consequently for disposal system design): Canister performance; Derived parameters of geosphere retention properties; Spatial variability of groundwater flow and geochemistry of migration paths; Time dependency (more emphasis on transient, time-dependent processes); Buffer evolution over long time. 488 refs.

  16. SKI SITE-94. Deep Repository Performance Assessment Project Volume I and II

    International Nuclear Information System (INIS)

    1996-12-01

    The function of SITE-94 is to provide the Swedish Nuclear Power Inspectorate (SKI) with the capacity and supporting knowledge needed for reviewing the Swedish nuclear industry's R and D programs and for reviewing license applications, as stipulated in Swedish legislation. The report is structured as a Performance Assessment exercise needed for input to decisions regarding repository safety, but the SITE-94 is not a safety assessment or a model for future assessments to be undertaken by the prospective licensee. The specific project objectives of SITE-94 comprise site evaluation, performance assessment methodology, canister integrity and radionuclide release and transport calculations. The report gives a detailed description of the many inter-related studies undertaken as part of the research project. As a general conclusion it is noted that the SITE-94 project has significantly advanced SKI's capability of reviewing performance assessments and has led to the development of a tool-kit for carrying out assessment calculations, both in terms of numerical and scoping analyses. The study has highlighted a number of key issues for safety assessment (and consequently for disposal system design): Canister performance; Derived parameters of geosphere retention properties; Spatial variability of groundwater flow and geochemistry of migration paths; Time dependency (more emphasis on transient, time-dependent processes); Buffer evolution over long time. 488 refs

  17. Investigation of modern methods of probalistic sensitivity analysis of final repository performance assessment models (MOSEL)

    International Nuclear Information System (INIS)

    Spiessl, Sabine; Becker, Dirk-Alexander

    2017-06-01

    Sensitivity analysis is a mathematical means for analysing the sensitivities of a computational model to variations of its input parameters. Thus, it is a tool for managing parameter uncertainties. It is often performed probabilistically as global sensitivity analysis, running the model a large number of times with different parameter value combinations. Going along with the increase of computer capabilities, global sensitivity analysis has been a field of mathematical research for some decades. In the field of final repository modelling, probabilistic analysis is regarded a key element of a modern safety case. An appropriate uncertainty and sensitivity analysis can help identify parameters that need further dedicated research to reduce the overall uncertainty, generally leads to better system understanding and can thus contribute to building confidence in the models. The purpose of the project described here was to systematically investigate different numerical and graphical techniques of sensitivity analysis with typical repository models, which produce a distinctly right-skewed and tailed output distribution and can exhibit a highly nonlinear, non-monotonic or even non-continuous behaviour. For the investigations presented here, three test models were defined that describe generic, but typical repository systems. A number of numerical and graphical sensitivity analysis methods were selected for investigation and, in part, modified or adapted. Different sampling methods were applied to produce various parameter samples of different sizes and many individual runs with the test models were performed. The results were evaluated with the different methods of sensitivity analysis. On this basis the methods were compared and assessed. This report gives an overview of the background and the applied methods. The results obtained for three typical test models are presented and explained; conclusions in view of practical applications are drawn. At the end, a recommendation

  18. Investigation of modern methods of probalistic sensitivity analysis of final repository performance assessment models (MOSEL)

    Energy Technology Data Exchange (ETDEWEB)

    Spiessl, Sabine; Becker, Dirk-Alexander

    2017-06-15

    Sensitivity analysis is a mathematical means for analysing the sensitivities of a computational model to variations of its input parameters. Thus, it is a tool for managing parameter uncertainties. It is often performed probabilistically as global sensitivity analysis, running the model a large number of times with different parameter value combinations. Going along with the increase of computer capabilities, global sensitivity analysis has been a field of mathematical research for some decades. In the field of final repository modelling, probabilistic analysis is regarded a key element of a modern safety case. An appropriate uncertainty and sensitivity analysis can help identify parameters that need further dedicated research to reduce the overall uncertainty, generally leads to better system understanding and can thus contribute to building confidence in the models. The purpose of the project described here was to systematically investigate different numerical and graphical techniques of sensitivity analysis with typical repository models, which produce a distinctly right-skewed and tailed output distribution and can exhibit a highly nonlinear, non-monotonic or even non-continuous behaviour. For the investigations presented here, three test models were defined that describe generic, but typical repository systems. A number of numerical and graphical sensitivity analysis methods were selected for investigation and, in part, modified or adapted. Different sampling methods were applied to produce various parameter samples of different sizes and many individual runs with the test models were performed. The results were evaluated with the different methods of sensitivity analysis. On this basis the methods were compared and assessed. This report gives an overview of the background and the applied methods. The results obtained for three typical test models are presented and explained; conclusions in view of practical applications are drawn. At the end, a recommendation

  19. Performance Assessment of a Generic Repository in Bedded Salt for DOE-Managed Nuclear Waste

    Science.gov (United States)

    Stein, E. R.; Sevougian, S. D.; Hammond, G. E.; Frederick, J. M.; Mariner, P. E.

    2016-12-01

    A mined repository in salt is one of the concepts under consideration for disposal of DOE-managed defense-related spent nuclear fuel (SNF) and high level waste (HLW). Bedded salt is a favorable medium for disposal of nuclear waste due to its low permeability, high thermal conductivity, and ability to self-heal. Sandia's Generic Disposal System Analysis framework is used to assess the ability of a generic repository in bedded salt to isolate radionuclides from the biosphere. The performance assessment considers multiple waste types of varying thermal load and radionuclide inventory, the engineered barrier system comprising the waste packages, backfill, and emplacement drifts, and the natural barrier system formed by a bedded salt deposit and the overlying sedimentary sequence (including an aquifer). The model simulates disposal of nearly the entire inventory of DOE-managed, defense-related SNF (excluding Naval SNF) and HLW in a half-symmetry domain containing approximately 6 million grid cells. Grid refinement captures the detail of 25,200 individual waste packages in 180 disposal panels, associated access halls, and 4 shafts connecting the land surface to the repository. Equations describing coupled heat and fluid flow and reactive transport are solved numerically with PFLOTRAN, a massively parallel flow and transport code. Simulated processes include heat conduction and convection, waste package failure, waste form dissolution, radioactive decay and ingrowth, sorption, solubility limits, advection, dispersion, and diffusion. Simulations are run to 1 million years, and radionuclide concentrations are observed within an aquifer at a point approximately 4 kilometers downgradient of the repository. The software package DAKOTA is used to sample likely ranges of input parameters including waste form dissolution rates and properties of engineered and natural materials in order to quantify uncertainty in predicted concentrations and sensitivity to input parameters. Sandia

  20. Spent fuel performance assessment (SPA) for a hypothetical repository in crystalline formations in Germany

    International Nuclear Information System (INIS)

    Luehrmann, L.; Noseck, U.; Storck, R.

    2000-07-01

    Within the framework of this project a first long-term safety assessment study for a generic German repository with spent nuclear fuel in granite host formations has been performed. Conceptual models have been developed and implemented into the numerical codes. These models describe the relevant processes in the near and far field of the repository. For the nuclide mobilisation and the diffusion-controlled transport through the bentonite barrier the computer code GRAPOS for far-field transport the code CHETMAD has been developed. Transport in the far field has been assumed to take place in a fracture network. As retardation mechanism matrix diffusion accompanied by linear equilibrium sorption on the rock matrix is considered. Both codes have been tested by intercomparison with codes of other countries. The dose rates have been calculated by the code EXCON considering the transport pathways into the biosphere. A reference scenario has been defined. It considers instantaneous saturation of the bentonite immediately after the operational phase of the repository, failure of all containers after 1000 years, diffusion through the bentonite, transport through fractured dykes, which represent a fast transport pathway in the low permeability region of the granite. The nuclide mobilization has been calculated according to a common source term which has been developed by all participants of the SPA project. It is assumed that 25% of the containers are connected to the considered transport pathway in the far field. The nuclides are transported to layers close to the surface. The contaminated water is pumped from a surface well and used for drinking, irrigation, cattle feed and fish ponds. (orig.) [de

  1. Development of models for use in the assessment of waste repository performance

    International Nuclear Information System (INIS)

    Dickson, A.G.; Weare, J.H.

    1989-09-01

    The work outlined in this proposal is intended both to provide thermodynamic data that is needed to assist in the assessment of waste repository performance and the modeling necessary to ascertain to what extent the data produced is consistent, both with itself and with other published data on related systems. During this stage of the research we shall endeavor to develop a model of the chemistry of aluminum in aqueous solution which is consistent with a wide variety of experimental data including data generated as part of this project together with data that has been published previously in the research literature. We propose a program of research designed to enable us to model the interaction of canister materials (e.g. copper and iron) with natural waters. Both experimental work and a modeling program are outlined. In the experimental program e.m.f. measurements and spectroscopic measurements will be made so as to determine the various association equilibria of iron and copper with the anions OH-, HCO 3 - , and CO 3 2- . The initial stages of the modeling program will concentrate on the identification and use of existing experimental data to produce a preliminary model. This will allow us to identify those areas where special emphasis should be placed to meet the needs of the waste disposal program objectives. The objective of this research is to produce thermodynamic data for use in the assessment of waste repository performance that has been measured using experimental procedures performed in accord with the Level 1 quality assurance requirements detailed in the L.L.N.L. Yucca Mountain Project Quality Procedures Manual. The modeling approach used in experimental planning and data assessment is a Level 3 activity. In addition to the establishment of the thermodynamic data base proposed here, results should lead to improved consistency in the overall modeling effort. 29 refs., 2 tabs

  2. Total System Performance Assessment, 1993: An evaluation of the potential Yucca Mountain repository

    International Nuclear Information System (INIS)

    Andrews, R.W.; Dale, T.F.; McNeish, J.A.

    1994-03-01

    Total System Performance Assessments are an important component in the evaluation of the suitability of Yucca Mountain, Nevada as a potential site for a mined geologic repository for the permanent disposal of high-level radioactive wastes in the United States. The Total System Performance Assessments are conducted iteratively during site characterization to identify issues which should be addressed by the characterization and design activities as well as providing input to regulatory/licensing and programmatic decisions. During fiscal years 1991 and 1992, the first iteration of Total System Performance Assessment (hereafter referred to as TSPA 1991) was completed by Sandia National Laboratories and Pacific Northwest Laboratory. Beginning in fiscal year 1993, the Civilian Radioactive Waste Management System Management and Operating Contractor was assigned the responsibility to plan, coordinate, and contribute to the second iteration of Total System Performance Assessment (hereafter referred to as TSPA 1993). This document presents the objectives, approach, assumptions, input, results, conclusions, and recommendations associated with the Management and Operating Contractor contribution to TSPA 1993. The new information incorporated in TSPA 1993 includes (1) revised estimates of radionuclide solubilities (and their thermal and geochemical dependency), (2) thermal and geochemical dependency of spent fuel waste alteration and glass dissolution rates, (3) new distribution coefficient (k d ) estimates, (4) revised estimates of gas-phase velocities and travel times, and (5) revised hydrologic modeling of the saturated zone which provides updated estimates of the advective flux through the saturated zone

  3. Total System Performance Assessment, 1993: An evaluation of the potential Yucca Mountain repository

    Energy Technology Data Exchange (ETDEWEB)

    Andrews, R.W.; Dale, T.F.; McNeish, J.A.

    1994-03-01

    Total System Performance Assessments are an important component in the evaluation of the suitability of Yucca Mountain, Nevada as a potential site for a mined geologic repository for the permanent disposal of high-level radioactive wastes in the United States. The Total System Performance Assessments are conducted iteratively during site characterization to identify issues which should be addressed by the characterization and design activities as well as providing input to regulatory/licensing and programmatic decisions. During fiscal years 1991 and 1992, the first iteration of Total System Performance Assessment (hereafter referred to as TSPA 1991) was completed by Sandia National Laboratories and Pacific Northwest Laboratory. Beginning in fiscal year 1993, the Civilian Radioactive Waste Management System Management and Operating Contractor was assigned the responsibility to plan, coordinate, and contribute to the second iteration of Total System Performance Assessment (hereafter referred to as TSPA 1993). This document presents the objectives, approach, assumptions, input, results, conclusions, and recommendations associated with the Management and Operating Contractor contribution to TSPA 1993. The new information incorporated in TSPA 1993 includes (1) revised estimates of radionuclide solubilities (and their thermal and geochemical dependency), (2) thermal and geochemical dependency of spent fuel waste alteration and glass dissolution rates, (3) new distribution coefficient (k{sub d}) estimates, (4) revised estimates of gas-phase velocities and travel times, and (5) revised hydrologic modeling of the saturated zone which provides updated estimates of the advective flux through the saturated zone.

  4. Abstraction of information in repository performance assessments. Examples from the SKI project Site-94

    International Nuclear Information System (INIS)

    Dverstorp, B.; Andersson, J.

    1995-01-01

    Performance Assessment of a nuclear waste repository implies an analysis of a complex system with many interacting processes. Even if some of these processes may be known to large detail, problems arise when combining all information, and means of abstracting information from complex detailed models into models that couple different processes are needed. Clearly, one of the major objectives of performance assessment, to calculate doses or other performance indicators, implies an enormous abstraction of information compared to all information that is used as input. Other problems are that the knowledge of different parts or processes is strongly variable and adjustments, interpretations, are needed when combining models from different disciplines. In addition, people as well as computers, even today, always have a limited capacity to process information and choices have to be made. However, because abstraction of information clearly is unavoidable in performance assessment the validity of choices made, always need to be scrutinized and judgements made need to be updated in an iterative process

  5. Development of an object-oriented simulation code for repository performance assessment

    International Nuclear Information System (INIS)

    Tsujimoto, Keiichi; Ahn, J.

    1999-01-01

    As understanding for mechanisms of radioactivity confinement by a deep geologic repository improves at the individual process level, it has become imperative to evaluate consequences of individual processes to the performance of the whole repository system. For this goal, the authors have developed a model for radionuclide transport in, and release from, the repository region by incorporating multiple-member decay chains and multiple waste canisters. A computer code has been developed with C++, an object-oriented language. By utilizing the feature that a geologic repository consists of thousands of objects of the same kind, such as the waste canister, the repository region is divided into multiple compartments and objects for simulation of radionuclide transport. Massive computational tasks are distributed over, and executed by, multiple networked workstations, with the help of parallel virtual machine (PVM) technology. Temporal change of the mass distribution of 28 radionuclides in the repository region for the time period of 100 million yr has been successfully obtained by the code

  6. SUMO, System performance assessment for a high-level nuclear waste repository: Mathematical models

    International Nuclear Information System (INIS)

    Eslinger, P.W.; Miley, T.B.; Engel, D.W.; Chamberlain, P.J. II.

    1992-09-01

    Following completion of the preliminary risk assessment of the potential Yucca Mountain Site by Pacific Northwest Laboratory (PNL) in 1988, the Office of Civilian Radioactive Waste Management (OCRWM) of the US Department of Energy (DOE) requested the Performance Assessment Scientific Support (PASS) Program at PNL to develop an integrated system model and computer code that provides performance and risk assessment analysis capabilities for a potential high-level nuclear waste repository. The system model that has been developed addresses the cumulative radionuclide release criteria established by the US Environmental Protection Agency (EPA) and estimates population risks in terms of dose to humans. The system model embodied in the SUMO (System Unsaturated Model) code will also allow benchmarking of other models being developed for the Yucca Mountain Project. The system model has three natural divisions: (1) source term, (2) far-field transport, and (3) dose to humans. This document gives a detailed description of the mathematics of each of these three divisions. Each of the governing equations employed is based on modeling assumptions that are widely accepted within the scientific community

  7. Regulatory compliance for a Yucca Mountain Repository: A performance assessment perspective

    International Nuclear Information System (INIS)

    Dyer, J.R.; Van Luik, A.E.; Gil, A.V.; Brocoum, S.J.

    1997-01-01

    The U.S. Department of Energy's Yucca Mountain Site Characterization Project is scheduled to submit a License Application in the year 2002. The License Application is to show compliance with the regulations promulgated by the U.S. Nuclear Regulatory Commission which implement standards promulgated by the U.S. Environmental Protection Agency. These standards are being revised, and it is not certain what their exact nature will be in term of either the performance measure(s) or the time frames that are to be addressed. This paper provides some insights pertaining to this regulatory history, an update on Yucca Mountain performance assessments, and a Yucca Mountain Site Characterization Project perspective on proper standards based on Project experience in performance assessment for its proposed Yucca Mountain Repository system. The Project's performance assessment based perspective on a proper standard applicable to Yucca Mountain may be summarized as follows: a proper standard should be straight forward and understandable; should be consistent with other standards and regulations; and should require a degree of proof that is scientifically supportable in a licensing setting. A proper standard should have several attributes: (1) propose a reasonable risk level as its basis, whatever the quantitative performance measure is chosen to be, (2) state a definite regulatory time frame for showing compliance with quantitative requirements, (3) explicitly recognize that the compliance calculations are not predictions of actual future risks, (4) define the biosphere to which risk needs to be calculated in such a way as to constrain potentially endless speculation about future societies and future human actions, and (5) have as its only quantitative requirement the risk limit (or surrogate performance measure keyed to risk) for the total system

  8. Achieving transparency in the total system performance assessment of a potential high level radioactive waste repository at Yucca Mountain, Nevada

    International Nuclear Information System (INIS)

    Bailey, J.; Rickertsen, L.; Cotton, T.

    1999-01-01

    This paper has presented an approach to quantitative assessment of the degree of postclosure defense-in-depth provided by the reference system of the VA (Viability Assessment). This approach identifies principal barriers, assesses barriers for common uncertainty and failure mode, conducts barrier neutralization analyses, and evaluates overall defense-in-depth. The neutralization approach of step 3 is particularly useful in untangling the contributions of various barriers to the results calculated by performance assessments. In fact, it provides the only way of assessing the contribution of barriers that are fully redundant with one another. The approach has been applied to the VA reference system. It shows how the natural transport barriers contribute to performance of the system. Since their individual contributions are redundant, uncertainties in those individual contributions are reduced in importance. The analyses also suggest uncertainties common to both of these barriers are important to the safety assessment. Thus, the approach appears to be capable of determining the contribution of the principal barriers to system performance. The ability to use performance assessment to show not only how the repository system is expected to perform, but also how it achieves that performance, should contribute substantially towards providing needed transparency to the safety case for a geologic repository. It is also a valuable tool during the development of the repository design and associated safety case, by identifying areas in which performance would be enhanced by increased redundancy

  9. Uncertainty and sensitivity analysis in performance assessment for the proposed high-level radioactive waste repository at Yucca Mountain, Nevada

    International Nuclear Information System (INIS)

    Helton, Jon C.; Hansen, Clifford W.; Sallaberry, Cédric J.

    2012-01-01

    Extensive work has been carried out by the U.S. Department of Energy (DOE) in the development of a proposed geologic repository at Yucca Mountain (YM), Nevada, for the disposal of high-level radioactive waste. As part of this development, a detailed performance assessment (PA) for the YM repository was completed in 2008 and supported a license application by the DOE to the U.S. Nuclear Regulatory Commission (NRC) for the construction of the YM repository. The following aspects of the 2008 YM PA are described in this presentation: (i) conceptual structure and computational organization, (ii) uncertainty and sensitivity analysis techniques in use, (iii) uncertainty and sensitivity analysis for physical processes, and (iv) uncertainty and sensitivity analysis for expected dose to the reasonably maximally exposed individual (RMEI) specified the NRC’s regulations for the YM repository. - Highlights: ► An overview of performance assessment for the proposed Yucca Mountain radioactive waste repository is presented. ► Conceptual structure and computational organization are described. ► Uncertainty and sensitivity analysis techniques are described. ► Uncertainty and sensitivity analysis results for physical processes are presented. ► Uncertainty and sensitivity analysis results for expected dose are presented.

  10. Viability Assessment of a Repository at Yucca Mountain. Volume 3: Total System Performance Assessment

    Energy Technology Data Exchange (ETDEWEB)

    None

    1998-12-01

    This volume reports the development of TSPA for the VA. This first section defines the general process involved in developing any TSPA, it describes the overall TSPA process as implemented by programs in the US and elsewhere in the world, and discusses the acceptability of TSPA as a process or tool for analyzing a nuclear waste repository system. Section 2 discusses the more specific use of the TSPA process for the TSPA-VA for Yucca Mountain, including goals, approach, and methods. It also includes a very brief synopsis of TSPA-VA results. Section 3 briefly discusses each of the component models that comprise the TSPA-VA. Each TSPA component model represents a discrete set of processes. The TSPA-VA components are: unsaturated zone flow, thermal hydrology, near- field geochemical environment, waste package degradation, waste form alteration and mobilization, unsaturated zone transport, saturated zone flow and transport, and biosphere. For each of these components, this section introduces the conceptualization of each individual process, describes the data sources, and discusses model parameter development and computer methods used to simulate each component. Section 4 explains the mechanics of how the individual TSPA components were combined into a ''base case'' and then provides the ''expected value'' results of a deterministic base case analysis. Section 4 also contains a description of the probabilistic analyses and results that help determine the relative importance of the various TSPA components and the data used to describe the components. Section 5 addresses sensitivity studies run for each of the TSPA components to understand how uncertainty in various parameters within a component change the TSPA results. Section 6 presents the findings of the sensitivity studies run on the various components in Section 5, and prioritizes the findings of the entire set of uncertainty and sensitivity studies of the components relative

  11. A Bayesian reliability approach to the performance assessment of a geological waste repository

    International Nuclear Information System (INIS)

    Flueck, J.A.; Singh, A.K.

    1992-01-01

    This paper discusses the task of selecting a suitable site for a high-level waste disposal repository (HLWR) which certainly is a complex one in that one must address both engineering and economic factors of the proposed facility and site as well as environmental, public health, safety, and sociopolitical factors. Acknowledging the complexity of the siting problem for a HLWR leads one to readily conclude that a formal analysis, including the use of a performance assessment model (PAM), is needed to assist the designated decision makers in their task of selecting a suitable site. The overall goal of a PAM is to aid the decision makers in making the best possible technical siting decision. For a number of reason, the authors believe that the combining of both Bayesian decision theory and reliability methodology provides the best approach to constructing a useful PAM for assisting in the siting of a HLWR. This combination allows one to formally integrate existing relevant information, professional judgement, and component model outputs to produce conditionally estimated probabilities for a decision tree approach to the radionuclide release problem of a proposed HLWR. If loss functions are available, this also allows one to calculate the expected costs or losses from possible radionuclide releases. This latter calculation may be very important in selecting the final site from among a number of alternative sites

  12. Sorption databases for the cementitious near-field of a L/ILW repository for performance assessment

    International Nuclear Information System (INIS)

    Bradbury, M.H.; Sarott, F.A.

    1995-03-01

    Approximately 95% of the material in the L/ILW repository for short-lived low- and intermediate-level wastes consists of concrete; the remaining approx. 5% consists of steel (4%) and high molecular weight organic waste components (1%). Radionuclide sorption onto concrete represents one of the most important retardation mechanisms in the disposal caverns. This report compiles the sorption properties of hydrated cement, the most important sorbing material present in concrete, in the form of data sets for safety relevant nuclides under repository conditions; these data can then be used directly in performance assessment. Processes which affect sorption onto cement in the disposal caverns are documented in different data sets in this report. In this report, the distribution coefficients for radionuclides on cement are based to a large extent on values measured under repository-relevant conditions; this is true for cement without complexants in particular. (author) figs., tabs., refs

  13. LISA: A performance assessment code for geological repositories of radioactive waste

    International Nuclear Information System (INIS)

    Bertozzi, G.; Saltelli, A.

    1985-01-01

    LISA, developed at JRC-Ispra, is a statistical code, which calculates the radiation exposures and risks associated with radionuclide releases from geological repositories of nuclear waste. The assessment methodology is described briefly. It requires that a number of probabilistic components be quantified and introduced in the analysis; the results are thus expressed in terms of risk. The subjective judgment of experts may be necessary to quantify the probabilities of occurrence of rare geological events. Because of large uncertainties in input data, statistical treatment of the Monte Carlo type is utilized for the analysis; thus, the output from LISA is obtained in the form of distributions. A few results of an application to a probabilistic scenario for a repository mined in a clay bed are illustrated

  14. Spent fuel as a waste form: Data needs to allow long term performance assessment under repository disposal conditions

    International Nuclear Information System (INIS)

    Oversby, V.M.

    1986-12-01

    Performance assessment calculations are required for high level waste repositories for a period of 10,000 years. The Siting Guidelines require a comparison of sites following site characterization and prior to final site selection to be made over a 100,000 year period. To perform the required calculations, a detailed knowledge of the physical and chemical processes that affect waste form performance will be needed for each site. This paper will review the factors that affect the release of radionuclides from spent fuel under repository conditions, summarize our present state of knowledge, and suggest areas where more work is needed to support the performance assessment calculations. 17 refs., 5 figs., 3 tabs

  15. A new approach to performance assessment of barriers in a repository. Executive summary, draft, technical appendices. Final report

    International Nuclear Information System (INIS)

    Mueller-Hoeppe, N.; Krone, J.; Niehues, N.; Poehler, M.; Raitz von Frentz, R.; Gauglitz, R.

    1999-06-01

    Multi-barrier systems are accepted as the basic approach for long term environmental safe isolation of radioactive waste in geological repositories. Assessing the performance of natural and engineered barriers is one of the major difficulties in producing evidence of environmental safety for any radioactive waste disposal facility, due to the enormous complexity of scenarios and uncertainties to be considered. This report outlines a new methodological approach originally developed basically for a repository in salt, but that can be transferred with minor modifications to any other host rock formation. The approach is based on the integration of following elements: (1) Implementation of a simple method and efficient criteria to assess and prove the tightness of geological and engineered barriers; (2) Using the method of Partial Safety Factors in order to assess barrier performance at certain reasonable level of confidence; (3) Integration of a diverse geochemical barrier in the near field of waste emplacement limiting systematically the radiological consequences from any radionuclide release in safety investigations and (4) Risk based approach for the assessment of radionuclide releases. Indicative calculations performed with extremely conservative assumptions allow to exclude any radiological health consequences from a HLW repository in salt to a reference person with a safety level of 99,9999% per year. (orig.)

  16. Iterative performance assessments as a regulatory tool for evaluating repository safety: How experiences from SKI Project-90 were used in formulating the new performance assessment project SITE-94

    International Nuclear Information System (INIS)

    Andersson, J.

    1993-01-01

    The Swedish Nuclear Power Inspectorate, SKI, regulatory research program has to prepare for the process of licensing a repository for spent nuclear fuel, by building up the necessary knowledge and review capacity. SKIs main strategy for meeting this demand is to develop an independent performance assessment capability. SKIs first own performance assessment project, Project-90, was completed in 1991 and is now followed by a new project, SITE-94. SITE-94 is based on conclusions reached within Project-90. An independent review of Project-90, carried out by a NEA team of experts, has also contributed to the formation of the project. Another important reason for the project is that the implementing organization in Sweden, SKB, has proposed to submit an application to start detailed investigation of a repository candidate site around 1997. SITE-94 is a performance assessment of a hypothetical repository at a real site. The main objective of the project is to determine how site specific data should be assimilated into the performance assessment process, and to evaluate how uncertainties inherent in site characterization will influence performance assessment results. This will be addressed by exploring multiple interpretations, conceptual models, and parameters consistent with the site data. The site evaluation will strive for consistency between geological, hydrological, rock mechanical, and geochemical descriptions. Other important elements of SITE-94 are the development of a practical and defensible methodology for defining, constructing and analyzing scenarios, the development of approaches for treatment of uncertainties, evaluation of canister integrity, and the development and application of an appropriate quality assurance plan for performance assessments

  17. 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)

  18. Methodology used for total system performance assessment of the potential nuclear waste repository at yucca mountain (USA)

    International Nuclear Information System (INIS)

    Devonec, E.; Sevougian, S.D.; Mattie, P.D.; Mcneish, J.A.; Mishra, S.

    2001-01-01

    The U.S. Department of Energy and its contractors are currently evaluating a site in Nevada (Yucca Mountain) for disposal of high-level radioactive waste from U.S. commercial nuclear plants and U.S. government-owned facilities. The suitability of the potential geologic repository is assessed, based on its performance in isolating the nuclear waste from the environment. Experimental data and models representing the natural and engineered barriers are combined into a Total System Performance Assessment (TSPA) model. Because of the uncertainty in the current data and in the future evolution of the total system, simulations follow a probabilistic approach. Multiple realization simulations using Monte Carlo analysis are conducted over time periods of up to one million years, which estimates a range of possible behaviors of the repository. In addition to the nominal scenario, other exposure scenarios include the possibility of disruptive events such as volcanic eruption or intrusion, or accidental human intrusion. Sensitivity to key uncertain processes is analyzed. The influence of stochastic variables on the TSPA model output is assessed by ''uncertainty importance analysis'', e.g., regression analysis and classification tree analysis. Further investigation of the impact of parameters and assumptions is conducted through ''one-off analysis'', which consists in fixing a parameter at a particular value, using an alternative conceptual model, or in making a different assumption. Finally, robustness analysis evaluates the performance of the repository when various natural or engineered barriers are assumed to be degraded. The objective of these analyses is to evaluate the performance of the potential repository system under conditions ranging from expected to highly unlikely, though physically possible conditions. (author)

  19. Methodology used for total system performance assessment of the potential nuclear waste repository at yucca mountain (USA)

    Energy Technology Data Exchange (ETDEWEB)

    Devonec, E.; Sevougian, S.D.; Mattie, P.D.; Mcneish, J.A. [Duke Engineering and Services, Town Center Drive, Las Vegas (United States); Mishra, S. [Duke Engineering and Services, Austin, TX (United States)

    2001-07-01

    The U.S. Department of Energy and its contractors are currently evaluating a site in Nevada (Yucca Mountain) for disposal of high-level radioactive waste from U.S. commercial nuclear plants and U.S. government-owned facilities. The suitability of the potential geologic repository is assessed, based on its performance in isolating the nuclear waste from the environment. Experimental data and models representing the natural and engineered barriers are combined into a Total System Performance Assessment (TSPA) model. Because of the uncertainty in the current data and in the future evolution of the total system, simulations follow a probabilistic approach. Multiple realization simulations using Monte Carlo analysis are conducted over time periods of up to one million years, which estimates a range of possible behaviors of the repository. In addition to the nominal scenario, other exposure scenarios include the possibility of disruptive events such as volcanic eruption or intrusion, or accidental human intrusion. Sensitivity to key uncertain processes is analyzed. The influence of stochastic variables on the TSPA model output is assessed by ''uncertainty importance analysis'', e.g., regression analysis and classification tree analysis. Further investigation of the impact of parameters and assumptions is conducted through ''one-off analysis'', which consists in fixing a parameter at a particular value, using an alternative conceptual model, or in making a different assumption. Finally, robustness analysis evaluates the performance of the repository when various natural or engineered barriers are assumed to be degraded. The objective of these analyses is to evaluate the performance of the potential repository system under conditions ranging from expected to highly unlikely, though physically possible conditions. (author)

  20. Privacy Impact Assessment (PIA) Repository

    Data.gov (United States)

    Department of Veterans Affairs — This repository contains Privacy Impact Assessments (PIA) that have been vetted/approved. Section 208 of the Electronic Government Act of 2002 (E-Gov Act) requires...

  1. Initial demonstration of the NRC`s capability to conduct a performance assessment for a High-Level Waste Repository

    Energy Technology Data Exchange (ETDEWEB)

    Codell, R.; Eisenberg, N.; Fehringer, D.; Ford, W.; Margulies, T.; McCartin, T.; Park, J.; Randall, J.

    1992-05-01

    In order to better review licensing submittals for a High-Level Waste Repository, the US Nuclear Regulatory Commission staff has expanded and improved its capability to conduct performance assessments. This report documents an initial demonstration of this capability. The demonstration made use of the limited data from Yucca Mountain, Nevada to investigate a small set of scenario classes. Models of release and transport of radionuclides from a repository via the groundwater and direct release pathways provided preliminary estimates of releases to the accessible environment for a 10,000 year simulation time. Latin hypercube sampling of input parameters was used to express results as distributions and to investigate model sensitivities. This methodology demonstration should not be interpreted as an estimate of performance of the proposed repository at Yucca Mountain, Nevada. By expanding and developing the NRC staff capability to conduct such analyses, NRC would be better able to conduct an independent technical review of the US Department of Energy (DOE) licensing submittals for a high-level waste (HLW) repository. These activities were divided initially into Phase 1 and Phase 2 activities. Additional phases may follow as part of a program of iterative performance assessment at the NRC. The NRC staff conducted Phase 1 activities primarily in CY 1989 with minimal participation from NRC contractors. The Phase 2 activities were to involve NRC contractors actively and to provide for the transfer of technology. The Phase 2 activities are scheduled to start in CY 1990, to allow Sandia National Laboratories to complete development and transfer of computer codes and the Center for Nuclear Waste Regulatory Analyses (CNWRA) to be in a position to assist in the acquisition of the codes.

  2. Review of scenario selection approaches for performance assessment of high-level waste repositories and related issues

    International Nuclear Information System (INIS)

    Banano, E.J.; Baca, R.G.

    1995-08-01

    The selection of scenarios representing plausible realizations of the future conditions-with associated probabilities of occurrence-that can affect the long-term performance of a high-level radioactive waste (HLW) repository is the commonly used method for treating the uncertainty in the prediction of the future states of the system. This method, conventionally referred to as the ''scenario approach,'' while common is not the only method to deal with this uncertainty; other method ''ch as the environmental simulation approach (ESA), have also been proposed. Two of the difficulties with the scenario approach are the lack of uniqueness in the definition of the term ''scenario'' and the lack of uniqueness in the approach to formulate scenarios, which relies considerably on subjective judgments. Consequently, it is difficult to assure that a complete and unique set of scenarios can be defined for use in a performance assessment. Because scenarios are key to the determination of the long-term performance of the repository system, this lack of uniqueness can present a considerable challenge when attempting to reconcile the set of scenarios, and their level of detail, obtained using different approaches, particularly among proponents and regulators of a HLW repository

  3. Review of scenario selection approaches for performance assessment of high-level waste repositories and related issues.

    Energy Technology Data Exchange (ETDEWEB)

    Banano, E.J. [Beta Corporation International, Albuquerque, NM (United States); Baca, R.G. [Southwest Research Inst., San Antonio, TX (United States). Center for Nuclear Waste Regulatory Analyses

    1995-08-01

    The selection of scenarios representing plausible realizations of the future conditions-with associated probabilities of occurrence-that can affect the long-term performance of a high-level radioactive waste (HLW) repository is the commonly used method for treating the uncertainty in the prediction of the future states of the system. This method, conventionally referred to as the ``scenario approach,`` while common is not the only method to deal with this uncertainty; other method ``ch as the environmental simulation approach (ESA), have also been proposed. Two of the difficulties with the scenario approach are the lack of uniqueness in the definition of the term ``scenario`` and the lack of uniqueness in the approach to formulate scenarios, which relies considerably on subjective judgments. Consequently, it is difficult to assure that a complete and unique set of scenarios can be defined for use in a performance assessment. Because scenarios are key to the determination of the long-term performance of the repository system, this lack of uniqueness can present a considerable challenge when attempting to reconcile the set of scenarios, and their level of detail, obtained using different approaches, particularly among proponents and regulators of a HLW repository.

  4. Radioactive waste isolation in salt: special advisory report on the status of the Office of Nuclear Waste Isolation's plans for repository performance assessment

    International Nuclear Information System (INIS)

    Ditmars, J.D.; Walbridge, E.W.; Rote, D.M.; Harrison, W.; Herzenberg, C.L.

    1983-10-01

    Repository performance assessment is analysis that identifies events and processes that might affect a repository system for isolation of radioactive waste, examines their effects on barriers to waste migration, and estimates the probabilities of their occurrence and their consequences. In 1983 Battelle Memorial Institute's Office of Nuclear Waste Isolation (ONWI) prepared two plans - one for performance assessment for a waste repository in salt and one for verification and validation of performance assessment technology. At the request of the US Department of Energy's Salt Repository Project Office (SRPO), Argonne National Laboratory reviewed those plans and prepared this report to advise SRPO of specific areas where ONWI's plans for performance assessment might be improved. This report presents a framework for repository performance assessment that clearly identifies the relationships among the disposal problems, the processes underlying the problems, the tools for assessment (computer codes), and the data. In particular, the relationships among important processes and 26 model codes available to ONWI are indicated. A common suggestion for computer code verification and validation is the need for specific and unambiguous documentation of the results of performance assessment activities. A major portion of this report consists of status summaries of 27 model codes indicated as potentially useful by ONWI. The code summaries focus on three main areas: (1) the code's purpose, capabilities, and limitations; (2) status of the elements of documentation and review essential for code verification and validation; and (3) proposed application of the code for performance assessment of salt repository systems. 15 references, 6 figures, 4 tables

  5. Uncertainty and sensitivity analysis for two-phase flow in the vicinity of the repository in the 1996 performance assessment for the Waste Isolation Pilot Plant: Disturbed conditions

    International Nuclear Information System (INIS)

    HELTON, JON CRAIG; BEAN, J.E.; ECONOMY, K.; GARNER, J.W.; MACKINNON, ROBERT J.; MILLER, JOEL D.; SCHREIBER, J.D.; VAUGHN, PALMER

    2000-01-01

    Uncertainty and sensitivity analysis results obtained in the 1996 performance assessment (PA) for the Waste Isolation Pilot Plant (WIPP) are presented for two-phase flow in the vicinity of the repository under disturbed conditions resulting from drilling intrusions. Techniques based on Latin hypercube sampling, examination of scatterplots, stepwise regression analysis, partial correlation analysis and rank transformations are used to investigate brine inflow, gas generation repository pressure, brine saturation and brine and gas outflow. Of the variables under study, repository pressure and brine flow from the repository to the Culebra Dolomite are potentially the most important in PA for the WIPP. Subsequent to a drilling intrusion repository pressure was dominated by borehole permeability and generally below the level (i.e., 8 MPa) that could potentially produce spallings and direct brine releases. Brine flow from the repository to the Culebra Dolomite tended to be small or nonexistent with its occurrence and size also dominated by borehole permeability

  6. Methodology Used for Total System Performance Assessment of the Potential Nuclear Waste Repository at Yucca Mountain (USA)

    International Nuclear Information System (INIS)

    E. Devibec; S.D. Sevougian; P.D. Mattie; J.A. McNeish; S. Mishra

    2001-01-01

    The U.S. Department of Energy and its contractors are currently evaluating a site in Nevada (Yucca Mountain) for disposal of high-level radioactive waste from U.S. commercial nuclear plants and U.S. government-owned facilities. The suitability of the potential geologic repository is assessed, based on its performance in isolating the nuclear waste from the environment. Experimental data and models representing the natural and engineered barriers are combined into a Total System Performance Assessment (TSPA) model [1]. Process models included in the TSPA model are unsaturated zone flow and transport, thermal hydrology, in-drift geochemistry, waste package degradation, waste form degradation, engineered barrier system transport, saturated zone flow and transport, and biosphere transport. Because of the uncertainty in the current data and in the future evolution of the total system, simulations follow a probabilistic approach. Multiple realization simulations using Monte Carlo analysis are conducted over time periods of up to one million years, which estimates a range of possible behaviors of the repository. The environmental impact is measured primarily by the annual dose received by an average member of a critical population group residing 20 km down-gradient of the potential repository. In addition to the nominal scenario, other exposure scenarios include the possibility of disruptive events such as volcanic eruption or intrusion, or accidental human intrusion. Sensitivity to key uncertain processes is analyzed. The influence of stochastic variables on the TSPA model output is assessed by ''uncertainty importance analysis'', e.g., regression analysis and classification tree analysis. Further investigation of the impact of parameters and assumptions is conducted through ''one-off analysis'', which consists in fixing a parameter at a particular value, using an alternative conceptual model, or in making a different assumption. Finally, robustness analysis evaluates

  7. Modeling The Inhalation Exposure Pathway In Performance Assessment Of Geologic Radioactive Waste Repository At Yucca Mountain

    International Nuclear Information System (INIS)

    M.A. Wasiolek

    2006-01-01

    Inhalation exposure pathway modeling has recently been investigated as one of the tasks of the BIOPROTA Project (BIOPROTA 2005). BIOPROTA was set up to address the key uncertainties in long term assessments of contaminant releases into the environment arising from radioactive waste disposal. Participants of this international Project include national authorities and agencies, both regulators and operators, with responsibility for achieving safe and acceptable radioactive waste management. The objective of the inhalation task was to investigate the calculation of doses arising from inhalation of particles suspended from soils within which long-lived radionuclides, particularly alpha emitters, had accumulated. It was recognized that site-specific conditions influence the choice of conceptual model and input parameter values. Therefore, one of the goals of the task was to identify the circumstances in which different processes included in specific inhalation exposure pathway models were important. This paper discusses evaluation of processes and modeling assumptions specific to the proposed repository at Yucca Mountain as compared to the typical approaches and other models developed for different assessments and project specific contexts. Inhalation of suspended particulates that originate from contaminated soil is an important exposure pathway, particularly for exposure to actinides such as uranium, neptunium and plutonium. Radionuclide accumulation in surface soil arises from irrigation of soil with contaminated water over many years. The level of radionuclide concentration in surface soil depends on the assumed duration of irrigation. Irrigation duration is one of the parameters used on biosphere models and it depends on a specific assessment context. It is one of the parameters addressed in this paper from the point of view of assessment context for the proposed repository at Yucca Mountain. The preferred model for the assessment of inhalation exposure uses

  8. The Use of Digital Repositories for Enhancing Teacher Pedagogical Performance

    Science.gov (United States)

    Cohen, Anat; Kalimi, Sharon; Nachmias, Rafi

    2013-01-01

    This research examines the usage of local learning material repositories at school, as well as related teachers' attitudes and training. The study investigates the use of these repositories for enhancing teacher performance and assesses whether the assimilation of the local repositories increases their usage of and contribution to by teachers. One…

  9. Implementation of computer codes for performance assessment of the Republic repository of LLW/ILW Mochovce

    International Nuclear Information System (INIS)

    Hanusik, V.; Kopcani, I.; Gedeon, M.

    2000-01-01

    This paper describes selection and adaptation of computer codes required to assess the effects of radionuclide release from Mochovce Radioactive Waste Disposal Facility. The paper also demonstrates how these codes can be integrated into performance assessment methodology. The considered codes include DUST-MS for source term release, MODFLOW for ground-water flow and BS for transport through biosphere and dose assessment. (author)

  10. Treatment of uncertainty in the US department of energy's yucca mountain repository total system performance assessment (TSPA)

    International Nuclear Information System (INIS)

    Van Luik, A.; Zwahlen, E.

    2004-01-01

    The regulatory requirements being addressed in the US geological repository programme for spent nuclear fuel and high-level waste specify that performance assessment is to be used to address probabilistically defined mean-value dose constraints. Dose was chosen as the preferred performance measure because an acceptable dose limit could be selected through the regulation-setting process, based on a defined acceptable risk. By setting a dose limit, arguments about the conversion of a potential dose to a potential risk was taken off the table as a potential licensing issue. However, the probabilistic approach called for actually delivers a 'risk of a dose', a risk of a potential given dose value to a hypothetical person living at a set distance from the repository, with a set lifestyle, between the time of permanent closure and 10 000 years. Analyses must also be shown for the peak dose if it occurs after 10 000 years, essentially to a million years. For uncertain parameters that are important to system performance, the goal is to present an analysis that, in accord with applicable regulation, focuses on the mean value of the performance measure but 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'. These disclosure requirements are to be met by showing a range of potential outcomes and designating the mean value within that range. (author)

  11. Repository Integration Program: RIP performance assessment and strategy evaluation model theory manual and user`s guide

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-11-01

    This report describes the theory and capabilities of RIP (Repository Integration Program). RIP is a powerful and flexible computational tool for carrying out probabilistic integrated total system performance assessments for geologic repositories. The primary purpose of RIP is to provide a management tool for guiding system design and site characterization. In addition, the performance assessment model (and the process of eliciting model input) can act as a mechanism for integrating the large amount of available information into a meaningful whole (in a sense, allowing one to keep the ``big picture`` and the ultimate aims of the project clearly in focus). Such an integration is useful both for project managers and project scientists. RIP is based on a `` top down`` approach to performance assessment that concentrates on the integration of the entire system, and utilizes relatively high-level descriptive models and parameters. The key point in the application of such a ``top down`` approach is that the simplified models and associated high-level parameters must incorporate an accurate representation of their uncertainty. RIP is designed in a very flexible manner such that details can be readily added to various components of the model without modifying the computer code. Uncertainty is also handled in a very flexible manner, and both parameter and model (process) uncertainty can be explicitly considered. Uncertainty is propagated through the integrated PA model using an enhanced Monte Carlo method. RIP must rely heavily on subjective assessment (expert opinion) for much of its input. The process of eliciting the high-level input parameters required for RIP is critical to its successful application. As a result, in order for any project to successfully apply a tool such as RIP, an enormous amount of communication and cooperation must exist between the data collectors, the process modelers, and the performance. assessment modelers.

  12. The Aespoe redox investigations in block scale. Project summary and implications for repository performance assessment

    Energy Technology Data Exchange (ETDEWEB)

    Banwart, S [ed.; Bradford Univ. (United Kingdom). Dept. of Civil Engineering

    1995-11-01

    During construction and operation of a deep repository for high level waste, the groundwater environment will be open to oxidizing surface conditions and surface water inflow. Construction of the access tunnel to the Hard Rock Laboratory intersected a vertical fracture zone at a depth of 70 m. Changes in the groundwater chemistry were then monitored for a period of three years. Mass balances for conservative and reactive natural tracers provided a basis for a site model of the fracture zone. Groundwater dilution was the dominant process affecting the geochemistry of the site. There was associated Na-Ca ion-exchange. There is evidence for extensive lateral flow between recharge areas and the discharge point in the access tunnel. There was a significant transport of young organic carbon from the surface to the discharge borehole during recharge. Proposed microbial processes during lateral flow include anaerobic respiration coupled to reduction of iron hydroxide, methanogenesis, precipitation of calcite and secondary ferrous minerals. There is strong evidence to rule out sulfate reduction. Sulfate increased dramatically, possibly by desorption from fracture minerals. Fracture performance is affected by the extensive groundwater dilution and by the microbial processes that occur. Dilution affects aqueous speciation, radionuclide solubility and sorption, and the stability and mobility of colloids. Aerobic respiration provides an effective sink for oxygen. Maintaining anoxia favors low solubility for Tc, Np, Pu and U. Reductive dissolution of iron hydroxide will release sorbed radionuclides, while formation of biogenic Fe(II) minerals can provide an effective reducing migration barrier to radionuclides and to future intrusion of oxygen from the surface. 26 refs, 13 figs, 10 tabs.

  13. The Aespoe redox investigations in block scale. Project summary and implications for repository performance assessment

    International Nuclear Information System (INIS)

    Banwart, S.

    1995-11-01

    During construction and operation of a deep repository for high level waste, the groundwater environment will be open to oxidizing surface conditions and surface water inflow. Construction of the access tunnel to the Hard Rock Laboratory intersected a vertical fracture zone at a depth of 70 m. Changes in the groundwater chemistry were then monitored for a period of three years. Mass balances for conservative and reactive natural tracers provided a basis for a site model of the fracture zone. Groundwater dilution was the dominant process affecting the geochemistry of the site. There was associated Na-Ca ion-exchange. There is evidence for extensive lateral flow between recharge areas and the discharge point in the access tunnel. There was a significant transport of young organic carbon from the surface to the discharge borehole during recharge. Proposed microbial processes during lateral flow include anaerobic respiration coupled to reduction of iron hydroxide, methanogenesis, precipitation of calcite and secondary ferrous minerals. There is strong evidence to rule out sulfate reduction. Sulfate increased dramatically, possibly by desorption from fracture minerals. Fracture performance is affected by the extensive groundwater dilution and by the microbial processes that occur. Dilution affects aqueous speciation, radionuclide solubility and sorption, and the stability and mobility of colloids. Aerobic respiration provides an effective sink for oxygen. Maintaining anoxia favors low solubility for Tc, Np, Pu and U. Reductive dissolution of iron hydroxide will release sorbed radionuclides, while formation of biogenic Fe(II) minerals can provide an effective reducing migration barrier to radionuclides and to future intrusion of oxygen from the surface. 26 refs, 13 figs, 10 tabs

  14. Total system performance assessment - 1995: An evaluation of the potential Yucca Mountain repository

    Energy Technology Data Exchange (ETDEWEB)

    Atkins, J.E.; Lee, J.H.; Lingineni, S.; Mishra, S; McNeish, J.A.; Sassani, D.C.; Sevougian, S.D.

    1995-11-01

    The U.S. Department of Energy (DOE) is currently investigating the feasibility of permanently disposing the nation`s commercial high-level radioactive wastes (in the form of spent fuel from the over 100 electric power-generating nuclear reactors across the U.S.) and a portion of the defense high-level radioactive wastes (currently stored at federal facilities around the country) in the unsaturated tuffaceous rocks at Yucca Mountain, Nevada. Quantitative predictions based on the most current understanding of the processes and parameters potentially affecting the long-term behavior of the disposal system are used to assess the ability of the site and its associated engineered designs to meet regulatory objectives of the US NRC and the US EPA. The evaluation of the ability of the overall system to meet the performance objectives specified in the applicable regulatory standards has been termed total system performance assessment (TSPA). Total system performance assessments require the explicit quantification of the relevant processes and process interactions. In addition assessments are useful to help define the most significant processes, the information gaps and uncertainties and therefore the additional information required for more robust and defensible assessment of the overall performance. The aim of any total system performance assessment is to be as complete and reasonably conservative as possible and to assure that the descriptions of the predictive models and parameters are sufficient to ascertain their accuracy. Total system performance assessments evolve with time. Previous iterations of total system performance assessment of the Yucca Mountain site and associated engineered barriers have been conducted in 1991 and 1993.

  15. Total system performance assessment - 1995: An evaluation of the potential Yucca Mountain repository

    International Nuclear Information System (INIS)

    Atkins, J.E.; Lee, J.H.; Lingineni, S.; Mishra, S.; McNeish, J.A.; Sassani, D.C.; Sevougian, S.D.

    1995-11-01

    The U.S. Department of Energy (DOE) is currently investigating the feasibility of permanently disposing the nation's commercial high-level radioactive wastes (in the form of spent fuel from the over 100 electric power-generating nuclear reactors across the U.S.) and a portion of the defense high-level radioactive wastes (currently stored at federal facilities around the country) in the unsaturated tuffaceous rocks at Yucca Mountain, Nevada. Quantitative predictions based on the most current understanding of the processes and parameters potentially affecting the long-term behavior of the disposal system are used to assess the ability of the site and its associated engineered designs to meet regulatory objectives of the US NRC and the US EPA. The evaluation of the ability of the overall system to meet the performance objectives specified in the applicable regulatory standards has been termed total system performance assessment (TSPA). Total system performance assessments require the explicit quantification of the relevant processes and process interactions. In addition assessments are useful to help define the most significant processes, the information gaps and uncertainties and therefore the additional information required for more robust and defensible assessment of the overall performance. The aim of any total system performance assessment is to be as complete and reasonably conservative as possible and to assure that the descriptions of the predictive models and parameters are sufficient to ascertain their accuracy. Total system performance assessments evolve with time. Previous iterations of total system performance assessment of the Yucca Mountain site and associated engineered barriers have been conducted in 1991 and 1993

  16. Total system performance assessment - 1995: An evaluation of the potential Yucca Mountain Repository

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-11-01

    The U.S. Department of Energy (DOE) is currently investigating the feasibility of permanently disposing the nation`s commercial high-level radioactive wastes (in the form of spent fuel from the over 100 electric power-generating nuclear reactors across the U.S.) and a portion of the defense high-level radioactive wastes (currently stored at federal facilities around the country) in the unsaturated tuffaceous rocks at Yucca Mountain, Nevada. Quantitative predictions based on the most current understanding of the processes and parameters potentially affecting the long-term behavior of the disposal system are used to assess the ability of the site and its associated engineered designs to meet regulatory objectives set forward by the U.S. Nuclear Regulatory Commission (NRC) and the U.S. Environmental Protection Agency (EPA). The evaluation of the ability of the overall system to meet the performance objectives specified in the applicable regulatory standards has been termed total system performance assessment (TSPA). The aim of any total system performance assessment is to be as complete and reasonably conservative as possible and to assure that the descriptions of the predictive models and parameters are sufficient to ascertain their accuracy. Total system performance assessments evolve with time. As additional site and design information is generated, performance assessment analyses can be revised to become more representative of the expected conditions and remove some of the conservative assumptions necessitated by the incompleteness of site and design data. Previous iterations of total system performance assessment of the Yucca Mountain site and associated engineered barriers have been conducted in 1991 and 1993. These analyses have been documented in Barnard, Eslinger, Wilson and Andrews.

  17. Total system performance assessment - 1995: An evaluation of the potential Yucca Mountain Repository

    International Nuclear Information System (INIS)

    1995-11-01

    The U.S. Department of Energy (DOE) is currently investigating the feasibility of permanently disposing the nation's commercial high-level radioactive wastes (in the form of spent fuel from the over 100 electric power-generating nuclear reactors across the U.S.) and a portion of the defense high-level radioactive wastes (currently stored at federal facilities around the country) in the unsaturated tuffaceous rocks at Yucca Mountain, Nevada. Quantitative predictions based on the most current understanding of the processes and parameters potentially affecting the long-term behavior of the disposal system are used to assess the ability of the site and its associated engineered designs to meet regulatory objectives set forward by the U.S. Nuclear Regulatory Commission (NRC) and the U.S. Environmental Protection Agency (EPA). The evaluation of the ability of the overall system to meet the performance objectives specified in the applicable regulatory standards has been termed total system performance assessment (TSPA). The aim of any total system performance assessment is to be as complete and reasonably conservative as possible and to assure that the descriptions of the predictive models and parameters are sufficient to ascertain their accuracy. Total system performance assessments evolve with time. As additional site and design information is generated, performance assessment analyses can be revised to become more representative of the expected conditions and remove some of the conservative assumptions necessitated by the incompleteness of site and design data. Previous iterations of total system performance assessment of the Yucca Mountain site and associated engineered barriers have been conducted in 1991 and 1993. These analyses have been documented in Barnard, Eslinger, Wilson and Andrews

  18. Sample performance assessment of a high-level radioactive waste repository: sensitivity analysis

    International Nuclear Information System (INIS)

    Tkaczyk, A.

    2001-01-01

    The Yucca Mountain Project (YMP) is the USA's first attempt at long-term storage of High-Level Radioactive Waste (HLW). In theory, the reasoning for such a repository seems sound. In practice, there are many scenarios and cases to be considered while putting such a project into effect. Since a goal of YMP is to minimize dangers associated with long-term storage of HLW, it is important to estimate the dose rate to which current and future generations will be subjected. The lifetime of the repository is simulated to indicate the radiation dose rate to the maximally exposed individual; it is assumed that if the maximally exposed individual would not be harmed by the annual dose, the remaining population will be at even smaller risk. The determination of what levels of exposure can be deemed harmless is a concern, and the results from the simulations as compared against various regulations are discussed. (author)

  19. Sample performance assessment of a high-level radioactive waste repository: sensitivity analysis

    Energy Technology Data Exchange (ETDEWEB)

    Tkaczyk, A. [Iowa State Univ. of Science and Technology, Ames, IA (United States). Dept. of Mechanical Engineering

    2001-07-01

    The Yucca Mountain Project (YMP) is the USA's first attempt at long-term storage of High-Level Radioactive Waste (HLW). In theory, the reasoning for such a repository seems sound. In practice, there are many scenarios and cases to be considered while putting such a project into effect. Since a goal of YMP is to minimize dangers associated with long-term storage of HLW, it is important to estimate the dose rate to which current and future generations will be subjected. The lifetime of the repository is simulated to indicate the radiation dose rate to the maximally exposed individual; it is assumed that if the maximally exposed individual would not be harmed by the annual dose, the remaining population will be at even smaller risk. The determination of what levels of exposure can be deemed harmless is a concern, and the results from the simulations as compared against various regulations are discussed. (author)

  20. Migration case studies and the implications of humic substances for the radiological performance assessment of radioactive waste repositories

    International Nuclear Information System (INIS)

    Bryan, N.D.; Bernhard, G.; Geipel, G.; Heise, K.H.; Schmeide, K.; Benes, P.

    2005-01-01

    Using the information obtained during the experimental and modelling tasks a series of migration case studies have been performed. These are not full performance assessment studies, but are merely intended to demonstrate the likely impact of humic substances upon the migration of radionuclides in the environment. Three separate sites have been investigated: (i) a shallow waste repository site at Dukovany in the Czech Republic; (ii) the formerly proposed, but now withdrawn, waste repository site at Gorleben in Germany; (iii) a large rock pile, consisting of tailings from a uranium mine, Schlema Alberode, near Dresden, Germany. To provide the maximum information, a variety of 'release scenarios' have been defined: continuous leakage in the Dukovany case study; a short pulsed release at Gorleben, and steady leaching in the case of the rock pile. The effects of the magnitudes of the chemical rate constants have been studied along with the effects of the initial distribution between exchangeable and non-exchangeable. Further, modelling during the HUPA project has shown that in certain lab column experiments the sorption of humic and humic-radionuclide complexes onto mineral surfaces could have an impact upon migration. Therefore, the likely impact of these processes on the field scale was investigated. In each of the three cases, humics are predicted to have a significant impact upon migration. However, whereas for the repository cases, it is the presence of the non-exchangeable fraction that results in migration, and the exchangeable fraction plays no significant part in transport, for the rock pile, the model predicts that the humic should promote migration even with the exchangeable interaction alone. Humic sorption may affect the degree of migration, but even in the case of maximum impact, the net effect of humic substances is still to enhance migration significantly. In fact, the initial distribution of radionuclide upon entry to the far-field is much more

  1. A methodology of uncertainty/sensitivity analysis for PA of HLW repository learned from 1996 WIPP performance assessment

    International Nuclear Information System (INIS)

    Lee, Y. M.; Kim, S. K.; Hwang, Y. S.; Kang, C. H.

    2002-01-01

    The WIPP (Waste Isolation Pilot Plant) is a mined repository constructed by the US DOE for the permanent disposal of transuranic (TRU) wastes generated by activities related to defence of the US since 1970. Its historical disposal operation began in March 1999 following receipt of a final permit from the State of NM after a positive certification decision for the WIPP was issued by the EPA in 1998, as the first licensed facility in the US for the deep geologic disposal of radioactive wastes. The CCA (Compliance Certification Application) for the WIPP that the DOE submitted to the EPA in 1966 was supported by an extensive Performance Assessment (PA) carried out by Sandia National Laboratories (SNL), with so-called 1996 PA. Even though such PA methodologies could be greatly different from the way we consider for HLW disposal in Korea largely due to quite different geologic formations in which repository are likely to be located, a review on lots of works done through the WIPP PA studies could be the most important lessons that we can learn from in view of current situation in Korea where an initial phase of conceptual studies on HLW disposal has been just started. The objective of this work is an overview of the methodology used in the recent WIPP PA to support the US DOE WIPP CCA ans a proposal for Korean case

  2. Development of models for use in the assessment of waste repository performance

    International Nuclear Information System (INIS)

    Dickson, A.G.; Weare, J.H.

    1989-01-01

    Any repository suitable for the disposal of radioactive waste comprises a combination of various man-made and natural barriers which aim to prevent transfer of the radionuclides from the canister to the external environment. Our group has developed the techniques needed to model the equilibrium chemistry of complex multicomponent rock/water systems over a range of ionic strengths and temperatures. These models, show remarkable agreement with field measurements when applied to natural systems. They have been applied successfully to interpret a variety of geochemical processes: studies of ancient and modern marine evaporite systems, studies of the solar evaporation of seawater studies of mineral formation in lakes and in fluid inclusions. At present our models are limited to the components of the seawater system (Na + , Mg 2+ , Ca 2+ , H + , SO 4 2- , HCO 3 - , CO 3 2- , CO 2 ), together with the borate species and SiO 2 . 2 tabs

  3. NRC iterative performance assessment Phase 2. Development of capabilities for review of a performance assessment for a high-level waste repository

    International Nuclear Information System (INIS)

    Wescott, R.G.; Lee, M.P.; Eisenberg, N.A.; Baca, R.G.

    1995-10-01

    In order to better review a potential license application to construct and operate a geologic repository for spent nuclear fuel and high-level radioactive waste (HLW), the Nuclear Regulatory Commission staff (and its contractor) has expanded and improved its capability to conduct performance assessments. This report documents the demonstration of the second phase of this capability. The demonstration made use of the scenario selection procedure developed by Sandia National Laboratories to provide a set of scenarios, with corresponding probabilities, for use in the consequence analysis of a potential HLW disposal site in unsaturated tuff. Models of release of radionuclides from the waste form and transport in ground water, air and by direct pathways provided preliminary estimates of releases to the accessible environment for a 10,000 year period. The input values of parameters necessary for the consequence models were sampled numerous times using Latin Hypercube Sampling from assumed probability distributions. The results from the consequence models were then used to generate Complementary Cumulative Distribution Functions (CCDFs) for either release to the accessible environment or effective dose equivalents to a target population. CCDFs were calculated for probabilistically significant combinations (scenarios) of four disruptive events; drilling, pluvial climate, seismicity and magmatism. Sensitivity and uncertainty analyses of the calculated releases and effective dose equivalents were also used to determine the importance of the parameters. Because of the preliminary nature of the analysis and the lack of an adequate data base, the results and conclusions presented in this report should be carefully interpreted. They should not be misconstrued to represent the actual performance of the proposed Yucca Mountain repository nor serve as an endorsement of the methods used

  4. Obtaining reasonable assurance on geochemical aspects of performance assessment of deep geologic repositories

    International Nuclear Information System (INIS)

    Van Luik, A.E.; Serne, R.J.

    1986-01-01

    Providing reasonable assurance that a deep geologic disposal system will perform as required by regulation involves, in part, the building of confidence by providing a sound scientific basis for the site characterization, engineered system design, and system performance modeling efforts. Geochemistry plays a role in each of these activities. Site characterization must result in a description of the in situ geochemical environment that will support the design of the engineered system and the modeling of the transport of specific radionuclides to the accessible environment. Judging the adequacy of this site characterization effort is a major aspect of providing reasonable assurance. Within site characterization, there are a number of geochemical issues that need to be addressed such as the usefulness of natural analog studies, and assessing the very long-term stability of the site geochemistry, given expected temperature and radiation conditions

  5. Groundwater age and lifetime expectancy modelling approach for site characterization and performance assessment of radwaste repository in clay formation

    International Nuclear Information System (INIS)

    Cornaton, F.; Perrochet, P.; Benabderrahmane, H.

    2010-01-01

    variables that are combined and used to assess the repository performance regarding its emplacement in the transposition zone. Age and lifetime expectancy distributions are solved considering advection and dispersion/diffusion processes according to the approach proposed by Cornaton and Perrochet (2006). Advective-dispersive age solutions are compared to available age dates of pore water within the two main calcareous aquifers (Dogger and Oxfordian) that embed the Callovo-Oxfordian host formation. Such a comparison is helpful for the consolidation of the flow calibration, the estimation of the transport porosity field (since porosity is age generator) and for analyzing the internal water mixing processes and hydraulic behavior of major faults. Lifetime expectancy solutions are used to predict the response at the biosphere resulting from contaminant mass input occurring at a series of hypothetical repository locations. Lifetime expectancy solutions combined with age solutions provide the distribution of total residence times within the domain (i.e the total time required to travel from recharge to discharge). The latter is used to map in the 3-D space the low- and high-speed flow zones at the local scale. Finally, the behavior of age solutions is investigated when the hydraulic regime is rendered transient in response to the climatic evolution during the past Million years and to the climatic projection for the coming Million years. Transient age solutions at actual time are useful to analyze the effect of the temporal flow regime variations on the measured age dates and on ages simulated under steady-flow conditions. (authors)

  6. Applicability of thermodynamic database of radioactive elements developed for the Japanese performance assessment of HLW repository

    International Nuclear Information System (INIS)

    Yui, Mikazu; Shibata, Masahiro; Rai, Dhanpat; Ochs, Michael

    2003-01-01

    In 1999 Japan Nuclear Cycle Development Institute (JNC) published a second progress report (also known as H12 report) on high-level radioactive waste (HLW) disposal in Japan (JNC 1999). This report helped to develop confidence in the selected HLW disposal system and to establish the implementation body in 2000 for the disposal of HLW. JNC developed an in-house thermodynamic database for radioactive elements for performance analysis of the engineered barrier system (EBS) and the geosphere for H12 report. This paper briefly presents the status of the JNC's thermodynamic database and its applicability to perform realistic analyses of the solubilities of radioactive elements, evolution of solubility-limiting solid phases, predictions of the redox state of Pu in the neutral pH range under reducing conditions, and to estimate solubilities of radioactive elements in cementitious conditions. (author)

  7. Validation of transport models for use in repository performance assessments: a view illustrated for INTRAVAL test case 1b

    International Nuclear Information System (INIS)

    Jackson, C.P.; Lever, D.A.; Sumner, P.J.

    1991-03-01

    We present our views on validation. We consider that validation is slightly different for general models and specific models. We stress the importance of presenting for review the case for (or against) a model. We outline a formal framework for validation, which helps to ensure that all the issues are addressed. Our framework includes calibration, testing predictions, comparison with alternative models, which we consider particularly important, analysis of discrepancies, presentation, consideration of implications and suggested improved experiments. We illustrate the approach by application to an INTRAVAL test case based on laboratory experiments. Three models were considered: a simple model that included the effects of advection, dispersion and equilibrium sorption, a model that also included the effects of rock-matrix diffusion, and a model with kinetic sorption. We show that the model with rock-matrix diffusion is the only one to provide a good description of the data. We stress the implications of extrapolating to larger length and time scales for repository performance assessments. (author)

  8. Distribution coefficient of radionuclides on rocks for performance assessment of high-level radioactive waste repository

    International Nuclear Information System (INIS)

    Shibutani, Tomoki; Shibata, Masahiro; Suyama, Tadahiro

    1999-11-01

    Distribution coefficients of radionuclides on rocks are selected for safety assessment in the 'Second Progress Report on Research and Development for the geological disposal of HLW in Japan (H12 Report)'. The categorized types of rock are granitic rocks (crystalline and acidic rocks), basaltic rocks (crystalline and basic rocks), psammitic rocks (neogene sedimentary (soft)), and tuffaceous-pelitic rocks (pre-neogene sedimentary rocks (hard)). The types of groundwater are FRHP (fresh reducing high-pH), FRLP (fresh reducing low-pH), SRHP (saline reducing high-pH), SRLP (saline reducing low-pH), MRNP (mixing reducing neutral-pH) and FOHP (fresh oxidizing high-pH) groundwater. The elements to be surveyed are Ni, Se, Zr, Nb, Tc, Pd, Sn, Cs, Sm, Pb, Ra, Ac, Th, Pa, U, Np, Pu, Am and Cm. Distribution coefficients are collected from literatures describing batch sorption experimental results, and are selected under consideration of conservativity. (author)

  9. Representation of two-phase flow in the vicinity of the repository in the 1996 performance assessment for the Waste Isolation Pilot Plant

    International Nuclear Information System (INIS)

    VAUGHN, PALMER; BEAN, J.E.; HELTON, JON CRAIG; LORD, MICHAEL E.; MACKINNON, ROBERT J.; SCHREIBER, JAMES D.

    2000-01-01

    The following topics related to the representation of two-phase (gas and brine) flow in the vicinity of the repository in the 1996 performance assessment (PA) for the Waste Isolation Pilot Plant (WIPP) are discussed: (1) system of nonlinear partial differential equations used to model two-phase flow, (2) incorporation of repository shafts into model (3) creep closure of repository. (4) interbed fracturing, (5) gas generation (6) capillary action in waste, (7) borebole model (8) numerical solution and (9) gas and brine flow across specified boundaries. Two-phase flow calculations are a central part of the 1996 WIPP PA and supply results that are subsequently used in the calculation of releases to the surface at the time of a drilling intrusion (i.e., spallings, direct brine releases) and long-term releases due to radionuclide transport by flowing groundwater

  10. Uncertainty and sensitivity analysis in the 2008 performance assessment for the proposed repository for high-level radioactive waste at Yucca Mountain, Nevada

    International Nuclear Information System (INIS)

    Helton, Jon Craig; Sallaberry, Cedric M.; Hansen, Clifford W.

    2010-01-01

    Extensive work has been carried out by the U.S. Department of Energy (DOE) in the development of a proposed geologic repository at Yucca Mountain (YM), Nevada, for the disposal of high-level radioactive waste. As part of this development, an extensive performance assessment (PA) for the YM repository was completed in 2008 (1) and supported a license application by the DOE to the U.S. Nuclear Regulatory Commission (NRC) for the construction of the YM repository (2). This presentation provides an overview of the conceptual and computational structure of the indicated PA (hereafter referred to as the 2008 YM PA) and the roles that uncertainty analysis and sensitivity analysis play in this structure.

  11. Nirex 97 an assessment of the post-closure performance of a deep waste repository at Sellafield. Volume 3; the groundwater pathway

    International Nuclear Information System (INIS)

    Baker, A.; Chambers, A.; Jackson, C.

    1997-01-01

    United Kingdom Nirex Limited ('Nirex') is responsible for providing and managing facilities for the safe disposal of intermediate and certain low-level radioactive waste (ILW and LLW respectively). Government policy is that the preferred disposal route for such wastes is a deep geological repository. The repository concept aims to use a combination of natural and engineered barriers to achieve the necessary degree of long-term isolation and containment of the radioactive wastes. Since 1987, Nirex has carried out an extensive technical programme directed at the science of safe disposal. The work comprises a research programme into the long-term performance of waste forms and the engineered and natural barriers, including the characterisation of candidate geological settings to assess their suitability to host a deep waste repository ('DWR'). Between mid-1991 and March 1997 the geological characterisation programme was concentrated on establishing the suitability, or otherwise, of a candidate site at Sellafield, West Cumbria. In July 1994, as part of a detailed site investigation programme, Nirex applied for planning permission to develop an underground Rock Characterisation Facility (RCF) at Longlands Farm near Sellafield. This application was rejected by the planning authority and Nirex's appeal against that decision led to a local planning inquiry which ran from September 1995 until February 1996. In line with the Inspector's Report, in March 1997 the Nirex appeal was dismissed by the Secretary of State for the Environment. The Company's response to that decision, and its readiness to contribute to the new government's review of the way forward, are described in the Nirex Annual Report for 1996-97. This report - Nirex 97 - is founded on the understanding developed through the Nirex technical programme. It reports the outcome of an assessment of the post-closure safety performance, over hundreds of thousands of years, of a repository system located in a potential

  12. Cementitious Near-Field Sorption Data Base for Performance Assessment of an ILW Repository in Opalinus Clay

    Energy Technology Data Exchange (ETDEWEB)

    Wieland, E.; Van Loon, L. R

    2003-08-01

    The present report describes a cement sorption database (SDB) for the safety-relevant radionuclides to be disposed of in the planned Swiss repository for long-lived intermediate-level radioactive wastes (ILW). This report is an update on earlier SDBs, which were compiled for the cementitious near field of a repository for low- and intermediate-level radioactive wastes (L/ILW) by BRADBURY + SAROTT (1995) and BRADBURY + VAN LOON (1998). The radionuclide inventories are determined by the waste streams to be disposed of in the ILW repository. A list of the safety-relevant radionuclides was provided based on the currently available information on ILW inventories. The compositions of the cement porewaters in the near fields of the L/ILW and ILW repositories, which had been calculated using well-established codes for modelling cement degradation, were compared to identify any differences in the near-field conditions and to assess their influence on radionuclide sorption. Sorption values were selected based on the previously reported SDBs for the near field of the L/ILW repository. Sorption values were revised if new information and/or data were available which allowed changes to or re-appraisals of the data to be made. The sorption values recommended in this report were either selected on the basis of data from in-house experimental studies or from literature data. For some key radioelements, i.e., Cs(l), Sr(II), Ni(II), Eu(lll), Th(IV) and Sn(IV), new data were available from in-house measurements. These elements had been selected for experimental studies due to their relevance to safety assessment and/or their importance as appropriate chemical analogues. Degradation products of bitumen and cellulose, concrete admixtures and cement-derived near-field colloids were taken into account as the main potential perturbations, which could reduce radionuclide sorption in the near field. Possible impacts of the perturbing factors on radionuclide mobility were considered and

  13. Spent fuel performance in geologic repository environments

    International Nuclear Information System (INIS)

    Bradley, D.J.

    1985-10-01

    The performance assessment of the waste package is a current area of study in the United States program to develop a geologic repository for nuclear waste isolation. The waste package is presently envisioned as the waste form and its surrounding containers and possibly a packing material composed of crushed host rock or mixtures of that rock with clays. This waste package is tied to performance criteria set forth in recent legislation. It is the goal of the Civilian Radioactive Waste Management Program to obtain the necessary information on the waste package, in several geologic environments, to show that the waste package provides reasonable assurance of meeting established performance criteria. This paper discusses the United States program directed toward managing high-level radioactive waste, with emphasis on the current effort to define the behavior of irradiated spent fuel in repository groundwaters. Current studies are directed toward understanding the rate and nature (such as valence state, colloid form if any, solid phase controlling solubility) of radionuclide release from the spent fuel. Due to the strong interactive effect of radiation, thermal fields, and waste package components on this release, current spent fuel studies are being conducted primarily in the presence of waste package components over a wide range of potential environments

  14. Nirex 97 an assessment of the post-closure performance of a deep waste repository at Sellafield. Volume 1: hydrogeological model development - conceptual basis and data

    International Nuclear Information System (INIS)

    Degnan, P.; Littleboy, A.

    1997-01-01

    United Kingdom Nirex Limited ('Nirex') is responsible for providing and managing facilities for the safe disposal of intermediate and certain low-level radioactive waste (ILW and LLW respectively). Government policy is that the preferred disposal route for such wastes is a deep geological repository. The repository concept aims to use a combination of natural and engineered barriers to achieve the necessary degree of long-term isolation and containment of the radioactive wastes. Since 1987, Nirex has carried out an extensive technical programme directed at the science of safe disposal. The work comprises a research programme into the long-term performance of waste forms and the engineered and natural barriers, including the characterisation of candidate geological settings to assess their suitability to host a deep waste repository ('DWR'). Between mid-1991 and March 1997 the geological characterisation programme was concentrated on establishing the suitability, or otherwise, of a candidate site at Sellafield, West Cumbria. In July 1994, as part of a detailed site investigation programme, Nirex applied for planning permission to develop an underground Rock Characterisation Facility (RCF) at Longlands Farm near Sellafield. This application was rejected by the planning authority and Nirex's appeal against that decision led to a local planning inquiry which ran from September 1995 until February 1996. In line with the Inspector's Report, in March 1997 the Nirex appeal was dismissed by the Secretary of State for the Environment. The Company's response to that decision, and its readiness to contribute to the new government's review of the way forward, are described in the Nirex Annual Report for 1996-97. This report - Nirex 97 - is founded on the understanding developed through the Nirex technical programme. It reports the outcome of an assessment of the post-closure safety performance, over hundreds of thousands of years, of a repository system located in a potential

  15. Generic performance assessment for a deep repository for low and intermediate level waste in the UK - a case study in assessing radiological impacts on the natural environment

    International Nuclear Information System (INIS)

    Jones, S.R.; Patton, D.; Copplestone D.; Norris, S.; O'Sullivan, P.

    2003-01-01

    Concentrations of radionuclides in soil and surface water, taken from a generic performance assessment of a repository for low and intermediate level radioactive waste, assumed to be located in the UK, have been used as the basis for a case study in assessing radiological impacts on the natural environment. Simplified descriptions of the terrestrial and aquatic ecosystem types likely to be impacted have been developed. A scoping assessment has identified 226 Ra, 210 Po, 234 U, 230 Th and 238 U as having the highest potential for impact, with doses from internally incorporated alpha emitters as being potentially of particular importance. These nuclides, together with 36 Cl and 129 I (which have proved to be of importance in radiological risk assessments for humans) were included in a more detailed dose assessment. A basic methodology for dose assessment of ecosystems is described, and has been applied for the defined impacted ecosystems. Paucity of published data on concentration factors prevented a more detailed assessment for terrestrial ecosystems. For the aquatic ecosystem, a more detailed assessment was possible and highest calculated absorbed dose rates (weighted for the likely higher biological effectiveness of alpha radiation were about 6.5 μGy h -1 . We conclude that harm to the impacted ecosystems is unlikely and make the observation that the lack of concentration factor or transfer factor data for a sufficiently wide range of species, ecosystems and nuclides appears to be the principal obstacle to establishing a comprehensive framework for the application of radiological protection to ecosystems

  16. Nirex 97 an assessment of the post-closure performance of a deep waste repository at Sellafield. Volume 2; hydrogeological conceptual model development - effective parameters and calibration appendix

    International Nuclear Information System (INIS)

    Jackson, C.; Watson, S.

    1997-01-01

    United Kingdom Nirex Limited ('Nirex') is responsible for providing and managing facilities for the safe disposal of intermediate and certain low-level radioactive waste (ILW and LLW respectively). Government policy is that the preferred disposal route for such wastes is a deep geological repository. The repository concept aims to use a combination of natural and engineered barriers to achieve the necessary degree of long-term isolation and containment of the radioactive wastes. Since 1987, Nirex has carried out an extensive technical programme directed at the science of safe disposal. The work comprises a research programme into the long-term performance of waste forms and the engineered and natural barriers, including the characterisation of candidate geological settings to assess their suitability to host a deep waste repository ('DWR'). Between mid-1991 and March 1997 the geological characterisation programme was concentrated on establishing the suitability, or otherwise, of a candidate site at Sellafield, West Cumbria. In July 1994, as part of a detailed site investigation programme, Nirex applied for planning permission to develop an underground Rock Characterisation Facility (RCF) at Longlands Farm near Sellafield. This application was rejected by the planning authority and Nirex's appeal against that decision led to a local planning inquiry which ran from September 1995 until February 1996. In line with the Inspector's Report, in March 1997 the Nirex appeal was dismissed by the Secretary of State for the Environment. The Company's response to that decision, and its readiness to contribute to the new government's review of the way forward, are described in the Nirex Annual Report for 1996-97. This report - Nirex 97 - is founded on the understanding developed through the Nirex technical programme. It reports the outcome of an assessment of the post-closure safety performance, over hundreds of thousands of years, of a repository system located in a potential

  17. Nirex 97 an assessment of the post-closure performance of a deep waste repository at Sellafield. Volume 2; hydrogeological conceptual model development - effective parameters and calibration

    International Nuclear Information System (INIS)

    Jackson, C.; Watson, S.

    1997-01-01

    United Kingdom Nirex Limited ('Nirex') is responsible for providing and managing facilities for the safe disposal of intermediate and certain low-level radioactive waste (ILW and LLW respectively). Government policy is that the preferred disposal route for such wastes is a deep geological repository. The repository concept aims to use a combination of natural and engineered barriers to achieve the necessary degree of long-term isolation and containment of the radioactive wastes. Since 1987, Nirex has carried out an extensive technical programme directed at the science of safe disposal. The work comprises a research programme into the long-term performance of waste forms and the engineered and natural barriers, including the characterisation of candidate geological settings to assess their suitability to host a deep waste repository ('DWR'). Between mid-1991 and March 1997 the geological characterisation programme was concentrated on establishing the suitability, or otherwise, of a candidate site at Sellafield, West Cumbria. In July 1994, as part of a detailed site investigation programme, Nirex applied for planning permission to develop an underground Rock Characterisation Facility (RCF) at Longlands Farm near Sellafield. This application was rejected by the planning authority and Nirex's appeal against that decision led to a local planning inquiry which ran from September 1995 until February 1996. In line with the Inspector's Report, in March 1997 the Nirex appeal was dismissed by the Secretary of State for the Environment. The Company's response to that decision, and its readiness to contribute to the new government's review of the way forward, are described in the Nirex Annual Report for 1996-97. This report - Nirex 97 - is founded on the understanding developed through the Nirex technical programme. It reports the outcome of an assessment of the post-closure safety performance, over hundreds of thousands of years, of a repository system located in a potential

  18. The role of the disturbed rock zone in radioactive waste repository safety and performance assessment. A topical discussion and international overview

    International Nuclear Information System (INIS)

    Winberg, A.

    1991-06-01

    A discussion was presented of the role and relative importance of the disturbed rock zone (DRZ) around the underground openings of a repository for nuclear waste in crystalline rock. The term disturbed rock zone was defined and possible criteria to be sued to distinguish if from undisturbed rock was suggested. The processes decisive for the hydraulic characteristics of the DRZ were discussed. With regard to the integral hydraulic characteristics of the DRZ, the effects of the excavation methodology, stress redistribution, thermal changes, chemical changes and backfill were discussed. A review of in-situ observations of the DRZ was provided. Model analysis where the DRZ has been explicitly or implicitly represented, either from a phenomenological and performance assessment aspect were reviewed. The implications of the disturbed rock zone for the safe performance of a nuclear waste repository were discussed. Conceptual models for the geometry of the DRZ and hydraulic conductivity distribution in the DRZ were suggested. (au) (82 refs.)

  19. Characterization, propagation and analysis of aleatory and epistemic uncertainty in the 2008 performance assessment for the proposed repository for radioactive waste at Yucca Mountain, Nevada

    International Nuclear Information System (INIS)

    Helton, Jon Craig; Sallaberry, Cedric M.; Hansen, Clifford W.

    2010-01-01

    The 2008 performance assessment (PA) for the proposed repository for high-level radioactive waste at Yucca Mountain (YM), Nevada, illustrates the conceptual structure of risk assessments for complex systems. The 2008 YM PA is based on the following three conceptual entities: a probability space that characterizes aleatory uncertainty; a function that predicts consequences for individual elements of the sample space for aleatory uncertainty; and a probability space that characterizes epistemic uncertainty. These entities and their use in the characterization, propagation and analysis of aleatory and epistemic uncertainty are described and illustrated with results from the 2008 YM PA.

  20. Important parameters in the performance of a potential repository at Yucca Mountain (TSPA-1995)

    International Nuclear Information System (INIS)

    Atkins, J.E.; Sevougian, S.D.; Lee, J.H.; Andrews, R.W.; McNeish, J.A.

    1996-01-01

    A total system performance assessment (TSPA) was conducted to determine how a potential repository at Yucca Mountain would behave. Using the results of this TSPA, regression was done to determine which parameters had the most important effect on the repository performance. These results were consistent with the current conceptual understanding of the repository

  1. Performance assessment

    International Nuclear Information System (INIS)

    Doe, T.

    1985-01-01

    The purpose of performance assessment is to show that the repository is expected to serve its stated function - disposing of radioactive waste safely both during operation and for the postclosure period. Performance assessment is a straightforward concept, but its application may be very complicated. The concept of performance assessment has been clarified by the Nuclear Regulatory Commission (NRC) in their Draft Generic Technical Position on Licensing Assessment Methodology for High-Level Waste Geologic Repositories (NRC, 1984). This document has gone a long way toward defining the criteria that the NRC will use to determine whether or not information from site characterization is adequate to meet the regulations of the Nuclear Regulatory Commission and the Environmental Protection Agency (EPA). A favorable determination is required for issuance of a construction authorization, which is the first major regulatory requirement for developing a working repository. It is, therefore, essential that a research program be developed that not only resolves the outstanding technical issues, but also does it in such a way that the results are clearly applicable to the formal performance assessment and licensing procedures. The definitions of performance assessment are reviewed and the current NRC thinking is summarized

  2. Expected dose for the early failure scenario classes in the 2008 performance assessment for the proposed high-level radioactive waste repository at Yucca Mountain, Nevada

    International Nuclear Information System (INIS)

    Helton, J.C.; Hansen, C.W.; Sallaberry, C.J.

    2014-01-01

    Extensive work has been carried out by the U.S. Department of Energy (DOE) in the development of a proposed geologic repository at Yucca Mountain (YM), Nevada, for the disposal of high-level radioactive waste. In support of this development and an associated license application to the U.S. Nuclear Regulatory Commission (NRC), the DOE completed an extensive performance assessment (PA) for the proposed YM repository in 2008. This presentation describes the determination of expected dose to the reasonably maximally exposed individual (RMEI) specified in the NRC regulations for the YM repository for the early waste package (WP) failure scenario class and the early drip shield (DS) failure scenario class in the 2008 YM PA. The following topics are addressed: (i) properties of the early failure scenario classes and the determination of dose and expected dose the RMEI, (ii) expected dose and uncertainty in expected dose to the RMEI from the early WP failure scenario class, (iii) expected dose and uncertainty in expected dose to the RMEI from the early DS failure scenario class, (iv) expected dose and uncertainty in expected dose to the RMEI from the combined early WP and early DS failure scenario class with and without the inclusion of failures resulting from nominal processes, and (v) uncertainty in the occurrence of early failure scenario classes. The present article is part of a special issue of Reliability Engineering and System Safety devoted to the 2008 YM PA; additional articles in the issue describe other aspects of the 2008 YM PA. - Highlights: • Extensive work has been carried out by the U.S. DOE in the development of a proposed geologic repository at Yucca Mountain (YM), Nevada, for the disposal of high-level radioactive waste. • Properties of the early failure scenario classes (i.e. early waste package failure and early drip shield failure) in the 2008 YM performance assessment are described. • Determination of dose, expected dose and expected (mean

  3. Repository environmental parameters and models/methodologies relevant to assessing the performance of high-level waste packages in basalt, tuff, and salt

    Energy Technology Data Exchange (ETDEWEB)

    Claiborne, H.C.; Croff, A.G.; Griess, J.C.; Smith, F.J.

    1987-09-01

    This document provides specifications for models/methodologies that could be employed in determining postclosure repository environmental parameters relevant to the performance of high-level waste packages for the Basalt Waste Isolation Project (BWIP) at Richland, Washington, the tuff at Yucca Mountain by the Nevada Test Site, and the bedded salt in Deaf Smith County, Texas. Guidance is provided on the identify of the relevant repository environmental parameters; the models/methodologies employed to determine the parameters, and the input data base for the models/methodologies. Supporting studies included are an analysis of potential waste package failure modes leading to identification of the relevant repository environmental parameters, an evaluation of the credible range of the repository environmental parameters, and a summary of the review of existing models/methodologies currently employed in determining repository environmental parameters relevant to waste package performance. 327 refs., 26 figs., 19 tabs.

  4. Repository environmental parameters and models/methodologies relevant to assessing the performance of high-level waste packages in basalt, tuff, and salt

    International Nuclear Information System (INIS)

    Claiborne, H.C.; Croff, A.G.; Griess, J.C.; Smith, F.J.

    1987-09-01

    This document provides specifications for models/methodologies that could be employed in determining postclosure repository environmental parameters relevant to the performance of high-level waste packages for the Basalt Waste Isolation Project (BWIP) at Richland, Washington, the tuff at Yucca Mountain by the Nevada Test Site, and the bedded salt in Deaf Smith County, Texas. Guidance is provided on the identify of the relevant repository environmental parameters; the models/methodologies employed to determine the parameters, and the input data base for the models/methodologies. Supporting studies included are an analysis of potential waste package failure modes leading to identification of the relevant repository environmental parameters, an evaluation of the credible range of the repository environmental parameters, and a summary of the review of existing models/methodologies currently employed in determining repository environmental parameters relevant to waste package performance. 327 refs., 26 figs., 19 tabs

  5. Use of simplified models in the performance assessment of a high-level waste repository system in Japan

    International Nuclear Information System (INIS)

    Pensado, Osvaldo; Mohanty, Sitakanta; Kanno, Takeshi; Tochigi, Yoshikatsu

    2005-01-01

    This paper explores simplifications to the H12 performance assessment model to enhance performance in Monte Carlo analyses. It is shown that similar reference case results to those of the H12 model can be derived by describing the buffer material surrounding a waste package as a planar body. Other possible simplifications to the performance assessment model in areas related to the stratification of the host rock transmissivity domain and solubility constraints in the buffer material are explored. (author)

  6. The Use of Numerical Models in Support of Site Characterization and Performance Assessment Studies of Geological Repositories. Results of an IAEA Coordinated Research Project 2005-2010

    International Nuclear Information System (INIS)

    2013-10-01

    The siting, development and operation of waste disposal facilities, and the related safety issues, have been described in many IAEA publications. The safe management and disposal of radioactive waste from the nuclear fuel cycle remains a necessary condition for future development of nuclear energy. In particular, the disposal of high level waste and spent nuclear fuel in geological repositories, despite having been studied worldwide over the past several decades, still requires full scale demonstration through safe implementation, as planned at the national level in Finland and Sweden by 2020 and 2023, respectively, and in France by 2025. Safety assessment techniques are currently applicable to potential facility location and development through a quite large range of approaches and methodologies. By implementing research activities through coordinated research projects (CRPs), the IAEA enables research institutes in both developing and developed Member States to collaborate on research topics of common interest. In response to requests by several Member States in different networks and platforms dealing with waste disposal, in 2005 a CRP on The Use of Numerical Models in Support of Site Characterization and Performance Assessment Studies of Geological Repositories was proposed and developed to transfer modelling expertise and numerical simulation technology to countries needing them for their national nuclear waste management programmes. All Member States involved in this CRP have acquired the scientific basis for, and expertise in, the site characterization process, including test design, data analysis, model calibration, model validation, predictive modelling, sensitivity analysis and uncertainty propagation analysis. This expertise is documented in this publication, in which numerical modelling is used to address the pertinent issue of site characterization and its impact on safety, using data and information from a potential repository site

  7. Alternative measure for performance of HLW geologic repository

    International Nuclear Information System (INIS)

    Joonhang, Ahn; Chambre, P.L.

    2001-01-01

    A repository performance model that can show effects of canister multiplicity and repository configuration has been developed. Masses of a radionuclide in the repository and in the far field are proposed as performance measures. Canister multiplicity has significant effects on the release of long-lived radionuclides from the repository. As more canisters are included in the same water stream, the radionuclide concentration in the stream increases, but becomes independent of the number of canisters for sufficiently many canisters. Effects of reduction of radionuclide mass in the repository on the repository performance are clearly observed if the canister multiplicity is taken into account and the mass-based measures are applied. (author)

  8. 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)

  9. Numerical modeling capabilities to predict repository performance

    International Nuclear Information System (INIS)

    1979-09-01

    This report presents a summary of current numerical modeling capabilities that are applicable to the design and performance evaluation of underground repositories for the storage of nuclear waste. The report includes codes that are available in-house, within Golder Associates and Lawrence Livermore Laboratories; as well as those that are generally available within the industry and universities. The first listing of programs are in-house codes in the subject areas of hydrology, solute transport, thermal and mechanical stress analysis, and structural geology. The second listing of programs are divided by subject into the following categories: site selection, structural geology, mine structural design, mine ventilation, hydrology, and mine design/construction/operation. These programs are not specifically designed for use in the design and evaluation of an underground repository for nuclear waste; but several or most of them may be so used

  10. The use of performance assessment for the potential high-level radioactive waste repository at Yucca Mountain, Nevada

    International Nuclear Information System (INIS)

    Lee, Joon H.; Andrews, R. W.

    1997-01-01

    This paper covers the introduction and overview of the Yucca Mountain site, the overview of waste package and EBS design, the organization of CRWMS M and O, the overview of total system performance assessment (TSPA), the components of TSPA model, the examples results of TSPA component models, and the example results of TSPA scoping sensitivity analyses. 22 figs

  11. Total System Performance Assessment, 1993: An evaluation of the potential Yucca Mountain repository, B00000000-01717-2200-00099, Rev. 01

    International Nuclear Information System (INIS)

    Andrews, R.W.; Dale, T.F.; McNeish, J.A.

    1994-03-01

    Total System Performance Assessments are an important component in the evaluation of the suitability of Yucca Mountain, Nevada as a potential site for a mined geologic repository for the permanent disposal of high-level radioactive wastes in the United States. The Total System Performance Assessments are conducted iteratively during the site characterization to identify issues which should be addressed by the characterization and design activities as well as providing input to regulatory/licensing and programmatic decisions. During fiscal years 1991 and 1992, the first iteration of Total System Performance Assessment (hereafter referred to as TSPA 1991) was completed by Sandia National Laboratories and Pacific Northwest Laboratory. Beginning in fiscal year 1993, the Civilian Radioactive Waste Management System Management and Operating Contractor was assigned the responsibility to plan, coordinate, and contribute to the second iteration of Total System Performance Assessment (hereafter referred to as TSPA 1993). This document presents the objectives, approach, assumptions, input, results, conclusions, and recommendations associated with the Management and Operating Contractor contribution to TSPA 1993. A parallel effort was conducted by Sandia National Laboratories and is reported in Wilson et al. (1994, in press)

  12. Methodology developed at the CEA/IPSN for logn term performance assessment of nuclear waste repositories in geological formations

    International Nuclear Information System (INIS)

    Raimbault, P.; Lewi, J.

    1985-05-01

    The CEA/ISPN is currently developing a methodology for safety evaluation of disposal site projects in granite, clay and bedded salt, host rocks formations. In the Institute of Protection and Nuclear Safety, the Department of Safety Analysis (DAS) is responsible for the coordination of the modeling effort which is performed in several specialized groups. The models are commissionned and utilized at the IPSN for specific safety evaluations. They are improved as needed and validated through international exercices (INTRACOIN-HYDROCOIN-ATKINS) and experimental programs. The DAS develops as well a global performance assessment code named MELODIE which structure allows to couple the individual models. This code participates to international joint studies such as PAGIS, in order to test its ability to model specific sites. This should help to control the adequation of the individual models to the risk assessment evaluation in order to insure the availability of specific data and to identify the most sensitive parameters. This approach should allow to coordinate the action between experimentation, code development and safety rules determination in order to be ready to perform safety assessment on chosen sites. The current status of the different aspects of this work is presented. The model development concerns mainly: transport, hydrogeology, source term, dose calculation and sensitivity studies. Its connection with data collection and model validation is stressed in the field of source modeling, hydrogeology, geochemistry and geoprospective. The description of the first version of MELODIE is presented. Some results of the interactive evaluation of the source term, the groundwater flow and the transport of radionuclides in a granite site are presented as well

  13. Electrochemical and micro-gravimetric corrosion studies on spent fuel provide relevant source term data for a repository performance assessment

    International Nuclear Information System (INIS)

    Wegen, Detlef H.; Bottomley, Paul D. W.; Glatz, Jean-Paul

    2004-01-01

    Various electrochemical methods (corrosion potential monitoring, AC impedance analysis and electrochemical noise monitoring) were used in the investigation of UO 2 samples: natural and doped with two different levels of 238 Pu (0.1 and 10 wt%) simulating the increasing α-intensities seen with time in the repository. The results were compared and were able to show the intense, but also the very local nature of the radiolysis and to demonstrate that corrosion rates were proportional to α-radiolysis and hence the 238 Pu content; the corrosion rates were in accordance with earlier work at ITU. By contrast it was seen that the redox potentials only gave information as to the bulk solution that did not reflect the true conditions at the electrode interface that were driving the corrosion processes of UO 2 dissolution in groundwaters. The study shows how electrochemical techniques can provide vital information on the corrosion mechanism at the UO 2 /solution interface

  14. Radioactive waste isolation in salt: peer review of the Office of Nuclear Waste Isolation's plans for repository performance assessment

    International Nuclear Information System (INIS)

    Harrison, W.

    1984-05-01

    In the peer review of the preliminary performance assessment plan, suggestions to improve the presentation are given, technical issues are addressed, a page-by-page commentary is included, and recommendations are summarized. In the peer review of the verification and validation plan, the following are addressed: suggestions to improve presentation; key activities that have been overlooked; improving the technical content of the plan; and degree to which code documents support the plan. Comments on other supporting code documents, page-by-page commentary, and summary of recommendations are included

  15. Development of a Risk-Based Performance Assessment Method for Long-Term Cover Systems--Application to the Monticello Mill Tailings Repository

    International Nuclear Information System (INIS)

    HO, CLIFFORD K.; ARNOLD, BILL W.; COCHRAN, JOHN R.; WEBB, STEPHEN W.; TAIRA, RANDAL Y.

    2001-01-01

    A probabilistic, risk-based performance-assessment methodology is being developed to assist designers, regulators, and involved stakeholders in the selection, design, and monitoring of long-term covers for contaminated subsurface sites. This report presents an example of the risk-based performance-assessment method using a repository site in Monticello, Utah. At the Monticello site, a long-term cover system is being used to isolate long-lived uranium mill tailings from the biosphere. Computer models were developed to simulate relevant features, events, and processes that include water flux through the cover, source-term release, vadose-zone transport, saturated-zone transport, gas transport, and exposure pathways. The component models were then integrated into a total-system performance-assessment model, and uncertainty distributions of important input parameters were constructed and sampled in a stochastic Monte Carlo analysis. Multiple realizations were simulated using the integrated model to produce cumulative distribution functions of the performance metrics, which were used to assess cover performance for both present- and long-term future conditions. Performance metrics for this study included the water percolation reaching the uranium mill tailings, radon flux at the surface, groundwater concentrations, and dose. Results of this study can be used to identify engineering and environmental parameters (e.g., liner properties, long-term precipitation, distribution coefficients) that require additional data to reduce uncertainty in the calculations and improve confidence in the model predictions. These results can also be used to evaluate alternative engineering designs and to identify parameters most important to long-term performance

  16. Approaches for treating uncertainty in the long term performance assessment of a geological waste repository in clay

    International Nuclear Information System (INIS)

    Marivoet, J.; Volckaert, G.; Wemaere, I.; Mallants, D.

    1998-01-01

    In Belgium the current strategy for high-level waste disposal is the geological disposal in a plastic clay layer. The performance assessment approach consists of a systematic scenario selection based on the FEP (features, events and processes) methodology followed by consequence analyses for the selected scenarios. In these consequence analyses the different sources of uncertainty are systematically considered. For the normal evolution scenario, i.e. the scenario which includes all FEP's which are about certain to occur, a stochastic technique of the Monte Carlo type is applied for treating uncertainty. For the altered evolution scenarios a deterministic approach is generally used to evaluate the uncertainties on the long-term. In the case of altered evolution scenarios, comparisons of fluxes from the far field into the biosphere with those calculated for the normal evolution scenario are used, beside dose calculations, to evaluate the safety consequences. Some typical examples of the above approaches will be presented. (author)

  17. Performance assessment for Nuclear Power Plant Krsko intermediate and low-level rad-waste repository; Provjera sigurnosnih performansi odlagalishta radioaktivnog otpada niske i srednje aktivnosti nuklearne elektrane Krsko

    Energy Technology Data Exchange (ETDEWEB)

    Jelavic, V [Inst. za Elektroprivredu, Zagreb. (Yugoslavia).; Tankosic, Dj [Bechtel Inc., San Francisco (United States); Skanata, D [Nuklearna elektrana Krshko, Krshko (Yugoslavia); Plecjas, I [Institut za Nuklearne Nauke Boris Kidric, Belgrade (Yugoslavia)

    1990-07-01

    Performance safety Assessment for NPP Krsko radwaste repository was performed (LLW/ILW). Shallow land and tunnel type concept were analyzed. Because it was based on two unknown referent sites, one for the shallow land concept and the other for the tunnel type, analysis was generic in nature. Scenario selecting process and consequence analysis were performed by using deterministic approach. Results for both concepts of disposal suggests that proposed NPP Krsko radwaste repository reference site and disposal technology will fully meet radiation limits imposed by the Yugoslav regulations and ICRP guidelines. (author)

  18. Determining redox properties of clay-rich sedimentary deposits in the context of performance assessment of radioactive waste repositories : Conceptual and practical aspects

    NARCIS (Netherlands)

    Behrends, T.; Bruggeman, Christophe

    Redox reactions play a key factor controlling the mobility of redox sensitive radionuclides in clay-rich sediments which might serve as host formations for radioactive waste repositories. Assessing the redox speciation of radionuclides requires information about the redox conditions in the formation

  19. Repository seal materials performance for a SALT Repository Project 5-year code/model development plan: Draft

    International Nuclear Information System (INIS)

    1987-06-01

    This document describes an integrated laboratory testing and model development effort for the seal system for a high-level nuclear waste repository in salt. The testing and modeling efforts are designed to determine seal material response in the repository environment, to provide models of seal system components for performance assessment, and to assist in the development of seal system designs. A code/model development and performance analysis program will be performed to predict the short- and long-term response of seal materials and seal components. The results from these analyses will be used to support the material testing activities on this contract and to support performance assessment activities that are conducted in other parts of the Salt Repository Project (SRP). 48 refs., 15 figs., 4 tabs

  20. Development of Earthquake Ground Motion Input for Preclosure Seismic Design and Postclosure Performance Assessment of a Geologic Repository at Yucca Mountain, NV

    International Nuclear Information System (INIS)

    I. Wong

    2004-01-01

    This report describes a site-response model and its implementation for developing earthquake ground motion input for preclosure seismic design and postclosure assessment of the proposed geologic repository at Yucca Mountain, Nevada. The model implements a random-vibration theory (RVT), one-dimensional (1D) equivalent-linear approach to calculate site response effects on ground motions. The model provides results in terms of spectral acceleration including peak ground acceleration, peak ground velocity, and dynamically-induced strains as a function of depth. In addition to documenting and validating this model for use in the Yucca Mountain Project, this report also describes the development of model inputs, implementation of the model, its results, and the development of earthquake time history inputs based on the model results. The purpose of the site-response ground motion model is to incorporate the effects on earthquake ground motions of (1) the approximately 300 m of rock above the emplacement levels beneath Yucca Mountain and (2) soil and rock beneath the site of the Surface Facilities Area. A previously performed probabilistic seismic hazard analysis (PSHA) (CRWMS M and O 1998a [DIRS 103731]) estimated ground motions at a reference rock outcrop for the Yucca Mountain site (Point A), but those results do not include these site response effects. Thus, the additional step of applying the site-response ground motion model is required to develop ground motion inputs that are used for preclosure and postclosure purposes

  1. Radionuclide transport in the vicinity of the repository and associated complementary cumulative distribution functions in the 1996 performance assessment for the Waste Isolation Pilot Plant

    International Nuclear Information System (INIS)

    Stockman, C.T.; Garner, J.W.; Helton, J.C.; Johnson, J.D.; Shinta, A.; Smith, L.N.

    2000-01-01

    The following topics related to radionuclide transport in the vicinity of the repository in the 1996 performance assessment for the Waste Isolation Pilot Plant are presented: (i) mathematical description of models; (ii) uncertainty and sensitivity analysis results arising from subjective (i.e. epistemic) uncertainty for individual releases; (iii) construction of complementary cumulative distribution functions (CCDFs) arising from stochastic (i.e. aleatory) uncertainty; and (iv) uncertainty and sensitivity analysis results for CCDFs. The presented results indicate that no releases to the accessible environment take place due to radionuclide movement through the anhydrite marker beds, through the Dewey Lake Red Beds or directly to the surface, and also that the releases to the Culebra Dolomite are small. Even when the effects of uncertain analysis inputs are taken into account, the CCDFs for release to the Culebra Dolomite fall to the left of the boundary line specified in the US Environmental Protection Agency's standard for the geologic disposal of radioactive waste (40 CFR 191, 40 CFR 194)

  2. Development of Earthquake Ground Motion Input for Preclosure Seismic Design and Postclosure Performance Assessment of a Geologic Repository at Yucca Mountain, NV

    Energy Technology Data Exchange (ETDEWEB)

    I. Wong

    2004-11-05

    This report describes a site-response model and its implementation for developing earthquake ground motion input for preclosure seismic design and postclosure assessment of the proposed geologic repository at Yucca Mountain, Nevada. The model implements a random-vibration theory (RVT), one-dimensional (1D) equivalent-linear approach to calculate site response effects on ground motions. The model provides results in terms of spectral acceleration including peak ground acceleration, peak ground velocity, and dynamically-induced strains as a function of depth. In addition to documenting and validating this model for use in the Yucca Mountain Project, this report also describes the development of model inputs, implementation of the model, its results, and the development of earthquake time history inputs based on the model results. The purpose of the site-response ground motion model is to incorporate the effects on earthquake ground motions of (1) the approximately 300 m of rock above the emplacement levels beneath Yucca Mountain and (2) soil and rock beneath the site of the Surface Facilities Area. A previously performed probabilistic seismic hazard analysis (PSHA) (CRWMS M&O 1998a [DIRS 103731]) estimated ground motions at a reference rock outcrop for the Yucca Mountain site (Point A), but those results do not include these site response effects. Thus, the additional step of applying the site-response ground motion model is required to develop ground motion inputs that are used for preclosure and postclosure purposes.

  3. 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

  4. Far-field sorption data bases for performance assessment of a L/ILW repository in an undisturbed Palfris marl host rock

    International Nuclear Information System (INIS)

    Bradbury, M.H.; Baeyens, B.

    1997-12-01

    A Palfris marl formation at Wellenberg (Gemeinde Wolfenschiessen, NW) has been chosen by NAGRA as a potential repository site for low- and intermediate-level radioactive waste, L/ILW. In the coming years a series of performance assessment studies will be performed for this site. One set of key data required for such safety analysis calculations is sorption data bases (SDB) for safety relevant radionuclides in the far-field. The purpose of this report is to describe the procedures used to generate sorption data bases appropriate for the in situ conditions existing along the different potential flow paths in an undisturbed marl host rock formation. An important aim was to document the sources of sorption data used and, in particular, the processes by which data selections were mad.e. The main guiding principles here were 'transparency' and 'traceability'. Inherent within this whole process is also the justification for, and defensibility of, the selected values. Much of the sorption data used to generate the SDB for marl came from the open literature. A major part of this report is concerned with describing the procedures whereby these initial literature values are modified so that they apply to the actual marl mineralogies and groundwater chemistries. The resulting 'reference R d values' are then further modified using so called Lab -> Field transfer factors to produce sorption values which are appropriate to the in situ bulk rock conditions. The Lab -> Field transfer factors attempt to correct for the differences in sorption site availability between the crushed rock state used in batch tests and the intact rock state existing in reality in the host rock. There are two main groundwater chemistries and five characteristic mineralogical compositions which cover the three broad types of flow paths which have been identified in the Palfris marl formation. In principle the methodology described here to construct sorption data bases for marl is applicable to any type of

  5. Far-field sorption data bases for performance assessment of a L/ILW repository in an undisturbed Palfris marl host rock

    Energy Technology Data Exchange (ETDEWEB)

    Bradbury, M.H.; Baeyens, B. [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

    1997-12-01

    A Palfris marl formation at Wellenberg (Gemeinde Wolfenschiessen, NW) has been chosen by NAGRA as a potential repository site for low- and intermediate-level radioactive waste, L/ILW. In the coming years a series of performance assessment studies will be performed for this site. One set of key data required for such safety analysis calculations is sorption data bases (SDB) for safety relevant radionuclides in the far-field. The purpose of this report is to describe the procedures used to generate sorption data bases appropriate for the in situ conditions existing along the different potential flow paths in an undisturbed marl host rock formation. An important aim was to document the sources of sorption data used and, in particular, the processes by which data selections were mad.e. The main guiding principles here were `transparency` and `traceability`. Inherent within this whole process is also the justification for, and defensibility of, the selected values. Much of the sorption data used to generate the SDB for marl came from the open literature. A major part of this report is concerned with describing the procedures whereby these initial literature values are modified so that they apply to the actual marl mineralogies and groundwater chemistries. The resulting `reference R{sub d} values` are then further modified using so called Lab -> Field transfer factors to produce sorption values which are appropriate to the in situ bulk rock conditions. The Lab -> Field transfer factors attempt to correct for the differences in sorption site availability between the crushed rock state used in batch tests and the intact rock state existing in reality in the host rock. There are two main groundwater chemistries and five characteristic mineralogical compositions which cover the three broad types of flow paths which have been identified in the Palfris marl formation. In principle the methodology described here to construct sorption data bases for marl is applicable to any

  6. Regulatory status on the safety assessment of a HLW repository in other countries

    International Nuclear Information System (INIS)

    Lee, Sung Ho; Hwang, Yong Soo

    2008-12-01

    To construct a HLW repository, it is essential to meet the requirements on the regulation for a deep geological disposal. Even if the construction of a HLW repository is determined positively, technical standards which assert the performance of a repository will be needed. Among various technical standards, safety assessment based on the repository evolution in the future will play an important role in the licensing process. The foreign countries' technical standards on the safety assessment of a HLW repository may be an indicator to carry out the R and D activities on geological disposal effectively. In this report, assessment period, limit of radiation dose and uncertainty related to the safety assessment are investigated and analyzed in detail. Especially, the technical reviews of USA regulation bodies seems to be reasonable in the point of the intrinsic attribute of safety assessment

  7. Assessment of cement durability in repository environment

    International Nuclear Information System (INIS)

    Ferreira, E.G.A.; Vicente, R.; Isiko, V.L.K.; Miyamoto, H.; Marumo, J.T.; Gobbo, L.A.

    2015-01-01

    The present research aimed at investigating the durability of cement paste under nuclear waste repository conditions using accelerated tests. Cement paste samples are examined after being exposed to the environmental conditions that are expected to prevail in the repository environment and the results are compared with those obtained with unexposed specimens or specimens exposed to reference conditions. The following exposure conditions were selected: a) Immersion in salt solution, distilled water, or kept in dry storage; b) Room temperature (20 C. degrees) or high temperature (60 C. degrees); c) Immersion time of 30 days or 60 days (not for dry storage); d) Irradiation to a dose of (400 kGy) or background radiation (0 kGy). After exposure to the stressing conditions, the effects of each factor on the cement paste samples were observed by changes in their characteristics. Compressive strength tests were performed on all samples and some of them were investigated in terms of changes in mineralogy by X-ray diffraction (XRD) and thermo-gravimetric analysis (TGA). With the results obtained so far it was possible to point out the following conclusions. First, after a period of immersion in water, cement paste samples further hydrated and presented higher mechanical resistance, as expected. Secondly, dry storage did not allow a complete hydration as a consequence of pore water evaporation. High temperatures intensified this process and led to the ettringite decomposition to meta-ettringite. Thirdly, higher temperature accelerated hydration kinetics and promoted higher mechanical resistance in samples kept under immersion. Fourthly, the irradiation dose applied was unable to change the mineralogy of cement paste samples and fifthly, no statistically significant differences were observed between 30 or 60 days exposure time, for the test conditions

  8. 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)

  9. Uncertainty management in radioactive waste repository site assessment

    International Nuclear Information System (INIS)

    Baldwin, J.f.; Martin, T.P.; Tocatlidou

    1994-01-01

    The problem of performance assessment of a site to serve as a repository for the final disposal of radioactive waste involves different types of uncertainties. Their main sources include the large temporal and spatial considerations over which safety of the system has to be ensured, our inability to completely understand and describe a very complex structure such as the repository system, lack of precision in the measured information etc. These issues underlie most of the problems faced when rigid probabilistic approaches are used. Nevertheless a framework is needed, that would allow for an optimal aggregation of the available knowledge and an efficient management of the various types of uncertainty involved. In this work a knowledge-based modelling of the repository selection process is proposed that through a consequence analysis, evaluates the potential impact that hypothetical scenarios will have on a candidate site. The model is organised around a hierarchical structure, relating the scenarios with the possible events and processes that characterise them, and the site parameters. The scheme provides for both crisp and fuzzy parameter values and uses fuzzy semantic unification and evidential support logic reference mechanisms. It is implemented using the artificial intelligence language FRIL and the interaction with the user is performed through a windows interface

  10. 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.

  11. 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

  12. Far Field Sorption Data Bases for Performance Assessment of a High-Level Radioactive Waste Repository in an Undisturbed Opalinus Clay Host Rock

    International Nuclear Information System (INIS)

    Bradburry, M.; Baeyens, B.

    2003-08-01

    An Opalinus Clay formation in the Zuercher Weinland is under consideration by Nagra as a potential location for a high-level and long-Iived intermediate-level radioactive waste repository. Performance assessment studies will be performed for this site and the purpose of this report is to describe the procedures used to develop sorption data bases appropriate for an undisturbed Opalinus Clay host rock which are required for such safety analysis calculations. In tight, low water content argillaceous rock formations such as Opalinus Clay, there is uncertainty concerning the in situ pH/P CO 2 . In order to take this intrinsic uncertainty into account porewater chemistries were calculated for a reference case, pH = 7.24, and for two other pH values, 6.3 and 7.8. Sorption data bases are given for the three cases. The basis for the sorption data bases is 'in-house' sorption measurements for Cs(I), Sr(II), Ni(II), Eu(III), Sn(IV), Se(IV), Th(IV) and I(-I) carried out on Opalinus Clay samples from Mont Terri (Canton Jura) since at the time the experiments were performed no core samples from the Benken borehole (Zuercher Weinland) were available. The Opalinus Clay at Mont Terri and Benken are part of the same geological formation . Despite having directly measured data for the above key radionuclides, some of the required distribution ratios (Rd) used to generate the sorption data bases still came from the open literature. An important part of this report is concerned with describing the procedures whereby these selected literature Rd values were modified so as to apply to the Benken Opalinus Clay mineralogy and groundwater chemistries calculated at the three pH values given above. The resulting Rd values were then further modified using so-called Lab→Field transfer factors to produce sorption values which were appropriate to the in situ bulk rock for the selected range of water chemistry conditions. Finally, it is important to have some appreciation of the uncertainties

  13. Far Field Sorption Data Bases for Performance Assessment of a High-Level Radioactive Waste Repository in an Undisturbed Opalinus Clay Host Rock

    Energy Technology Data Exchange (ETDEWEB)

    Bradburry, M.; Baeyens, B

    2003-08-01

    An Opalinus Clay formation in the Zuercher Weinland is under consideration by Nagra as a potential location for a high-level and long-Iived intermediate-level radioactive waste repository. Performance assessment studies will be performed for this site and the purpose of this report is to describe the procedures used to develop sorption data bases appropriate for an undisturbed Opalinus Clay host rock which are required for such safety analysis calculations. In tight, low water content argillaceous rock formations such as Opalinus Clay, there is uncertainty concerning the in situ pH/P{sub CO{sub 2}}. In order to take this intrinsic uncertainty into account porewater chemistries were calculated for a reference case, pH = 7.24, and for two other pH values, 6.3 and 7.8. Sorption data bases are given for the three cases. The basis for the sorption data bases is 'in-house' sorption measurements for Cs(I), Sr(II), Ni(II), Eu(III), Sn(IV), Se(IV), Th(IV) and I(-I) carried out on Opalinus Clay samples from Mont Terri (Canton Jura) since at the time the experiments were performed no core samples from the Benken borehole (Zuercher Weinland) were available. The Opalinus Clay at Mont Terri and Benken are part of the same geological formation . Despite having directly measured data for the above key radionuclides, some of the required distribution ratios (Rd) used to generate the sorption data bases still came from the open literature. An important part of this report is concerned with describing the procedures whereby these selected literature Rd values were modified so as to apply to the Benken Opalinus Clay mineralogy and groundwater chemistries calculated at the three pH values given above. The resulting Rd values were then further modified using so-called Lab{yields}Field transfer factors to produce sorption values which were appropriate to the in situ bulk rock for the selected range of water chemistry conditions. Finally, it is important to have some

  14. Brine and Gas Flow Patterns Between Excavated Areas and Disturbed Rock Zone in the 1996 Performance Assessment for the Waste Isolation Pilot Plant for a Single Drilling Intrusion that Penetrates Repository and Castile Brine Reservoir

    International Nuclear Information System (INIS)

    Economy, Kathleen M.; Helton, Jon Craig; Vaughn, Palmer

    1999-01-01

    The Waste Isolation Pilot Plant (WIPP), which is located in southeastern New Mexico, is being developed for the geologic disposal of transuranic (TRU) waste by the U.S. Department of Energy (DOE). Waste disposal will take place in panels excavated in a bedded salt formation approximately 2000 ft (610 m) below the land surface. The BRAGFLO computer program which solves a system of nonlinear partial differential equations for two-phase flow, was used to investigate brine and gas flow patterns in the vicinity of the repository for the 1996 WIPP performance assessment (PA). The present study examines the implications of modeling assumptions used in conjunction with BRAGFLO in the 1996 WIPP PA that affect brine and gas flow patterns involving two waste regions in the repository (i.e., a single waste panel and the remaining nine waste panels), a disturbed rock zone (DRZ) that lies just above and below these two regions, and a borehole that penetrates the single waste panel and a brine pocket below this panel. The two waste regions are separated by a panel closure. The following insights were obtained from this study. First, the impediment to flow between the two waste regions provided by the panel closure model is reduced due to the permeable and areally extensive nature of the DRZ adopted in the 1996 WIPP PA, which results in the DRZ becoming an effective pathway for gas and brine movement around the panel closures and thus between the two waste regions. Brine and gas flow between the two waste regions via the DRZ causes pressures between the two to equilibrate rapidly, with the result that processes in the intruded waste panel are not isolated from the rest of the repository. Second, the connection between intruded and unintruded waste panels provided by the DRZ increases the time required for repository pressures to equilibrate with the overlying and/or underlying units subsequent to a drilling intrusion. Third, the large and areally extensive DRZ void volumes is a

  15. Brine and Gas Flow Patterns Between Excavated Areas and Disturbed Rock Zone in the 1996 Performance Assessment for the Waste Isolation Pilot Plant for a Single Drilling Intrusion that Penetrates Repository and Castile Brine Reservoir

    Energy Technology Data Exchange (ETDEWEB)

    ECONOMY,KATHLEEN M.; HELTON,JON CRAIG; VAUGHN,PALMER

    1999-10-01

    The Waste Isolation Pilot Plant (WIPP), which is located in southeastern New Mexico, is being developed for the geologic disposal of transuranic (TRU) waste by the U.S. Department of Energy (DOE). Waste disposal will take place in panels excavated in a bedded salt formation approximately 2000 ft (610 m) below the land surface. The BRAGFLO computer program which solves a system of nonlinear partial differential equations for two-phase flow, was used to investigate brine and gas flow patterns in the vicinity of the repository for the 1996 WIPP performance assessment (PA). The present study examines the implications of modeling assumptions used in conjunction with BRAGFLO in the 1996 WIPP PA that affect brine and gas flow patterns involving two waste regions in the repository (i.e., a single waste panel and the remaining nine waste panels), a disturbed rock zone (DRZ) that lies just above and below these two regions, and a borehole that penetrates the single waste panel and a brine pocket below this panel. The two waste regions are separated by a panel closure. The following insights were obtained from this study. First, the impediment to flow between the two waste regions provided by the panel closure model is reduced due to the permeable and areally extensive nature of the DRZ adopted in the 1996 WIPP PA, which results in the DRZ becoming an effective pathway for gas and brine movement around the panel closures and thus between the two waste regions. Brine and gas flow between the two waste regions via the DRZ causes pressures between the two to equilibrate rapidly, with the result that processes in the intruded waste panel are not isolated from the rest of the repository. Second, the connection between intruded and unintruded waste panels provided by the DRZ increases the time required for repository pressures to equilibrate with the overlying and/or underlying units subsequent to a drilling intrusion. Third, the large and areally extensive DRZ void volumes is a

  16. Near Field sorption Data Bases for Compacted MX-80 Bentonite for Performance Assessment of a High-Level Radioactive Waste Repository in Opalinus Clay Host Rock

    Energy Technology Data Exchange (ETDEWEB)

    Bradbury, M.; Baeyens, B

    2003-08-01

    Bentonites of various types and compacted forms are being investigated in many countries as backfill materials in high-level radioactive waste disposal concepts. Nagra is currently considering an Opalinus clay (OPA) formation in the Zuercher Weinland as a potential location for a high-level radioactive waste repository. A compacted MX-80 bentonite is foreseen as a potential backfill material. Performance assessment studies will be performed for this site and one of the requirements for such an assessment are sorption data bases (SDB) for the bentonite near-field. The purpose of this report is to describe the procedures used to develop the SDB. One of the pre-requisites for developing a SDB is a water chemistry for the compacted bentonite porewater. For a number of reasons mentioned in the report, and discussed in more detail elsewhere, this is not a straightforward task. There are considerable uncertainties associated with the major ion concentrations and in particular with the system pH and Eh. The MX-80 SDB was developed for a reference bentonite porewater (pH = 7.25) which was calculated using the reference OPA porewater. In addition, two further SDBs are presented for porewaters calculated at pH values of 6.9 and 7.9 corresponding to lower and upper bound values calculated for the range of groundwater compositions anticipated for the OPA host rock. 'In house' sorption isotherm data were measured for Cs(I), Ni(II), Eu(III), Th(IV), Se(IV) and 1(-1) on the 'as received' MX-80 material equilibrated with a simulated porewater composition. Complementary 'in house' sorption edge and isotherm measurements on conditioned Na/Ca montmorillonites were also available for many of these radionuclides. These data formed the core of the SDB. Nevertheless, some of the required sorption data still had to be obtained from the open literature. An important part of this report is concerned with describing selection procedures and the modifications

  17. Review of Y/OWI/TM-36: repository design performance in salt, granite, shale or basalt

    International Nuclear Information System (INIS)

    Talbot, R.; Nair, O.B.

    1979-09-01

    As part of the ongoing work by the Lawrence Livermore Laboratory to evaluate repository design performance, this memorandum presents a review of the preconceptual repository design described in Y/OWI/TM-36, Technical Support for GEIS: Radioactive Waste Isolation in Geologic Formations, April 1978. The purpose of this review is: to assess the adequacy of the design procedures and assumptions; to identify inappropriate or unsubstantiated design issues; to identify areas where additional numerical analyses may be required; and to develop data for inclusion in a reference repository design. The preconceptual repository design is presented in the form of 23 volumes of data base, analyses, and design layouts for four rock types: bedded salt, shale, granite and basalt. This memorandum reviews all four repository designs

  18. Impact of Drill and Blast Excavation on Repository Performance Confirmation

    International Nuclear Information System (INIS)

    Keller, R.; Francis, N.; Houseworth, J.; Kramer, N.

    2000-01-01

    There has been considerable work accomplished internationally examining the effects of drill and blast excavation on rock masses surrounding emplacement openings of proposed nuclear waste repositories. However, there has been limited discussion tying the previous work to performance confirmation models such as those proposed for Yucca Mountain, Nevada. This paper addresses a possible approach to joining the available information on drill and blast excavation and performance confirmation. The method for coupling rock damage data from drill and blast models to performance assessment models for fracture flow requires a correlation representing the functional relationship between the peak particle velocity (PPV) vibration levels and the potential properties that govern water flow rates in the host rock. Fracture aperture and frequency are the rock properties which may be most influenced by drill and blast induced vibration. If it can be shown (using an appropriate blasting model simulation) that the effect of blasting is far removed from the waste package in an emplacement drift, then disturbance to the host rock induced in the process of drill and blast excavation may be reasonably ignored in performance assessment calculations. This paper proposes that the CANMET (Canada Center for Mineral and Energy Technology) Criterion, based on properties that determine rock strength, may be used to define a minimum PPV. This PPV can be used to delineate the extent of blast induced damage. Initial applications have demonstrated that blasting models can successfully be coupled with this criterion to predict blast damage surrounding underground openings. The Exploratory Studies Facility at Yucca Mountain has used a blasting model to generate meaningful estimates of near-field vibration levels and damage envelopes correlating to data collected from pre-existing studies conducted. Further work is underway to expand this application over a statistical distribution of geologic

  19. Proposed methodology for completion of scenario analysis for the Basalt Waste Isolation Project. [Assessment of post-closure performance for a proposed repository for high-level nuclear waste

    Energy Technology Data Exchange (ETDEWEB)

    Roberds, W.J.; Plum, R.J.; Visca, P.J.

    1984-11-01

    This report presents the methodology to complete an assessment of postclosure performance, considering all credible scenarios, including the nominal case, for a proposed repository for high-level nuclear waste at the Hanford Site, Washington State. The methodology consists of defensible techniques for identifying and screening scenarios, and for then assessing the risks associated with each. The results of the scenario analysis are used to comprehensively determine system performance and/or risk for evaluation of compliance with postclosure performance criteria (10 CFR 60 and 40 CFR 191). In addition to describing the proposed methodology, this report reviews available methodologies for scenario analysis, discusses pertinent performance assessment and uncertainty concepts, advises how to implement the methodology (including the organizational requirements and a description of tasks) and recommends how to use the methodology in guiding future site characterization, analysis, and engineered subsystem design work. 36 refs., 24 figs., 1 tab.

  20. Reliable predictions of waste performance in a geologic repository

    International Nuclear Information System (INIS)

    Pigford, T.H.; Chambre, P.L.

    1985-08-01

    Establishing reliable estimates of long-term performance of a waste repository requires emphasis upon valid theories to predict performance. Predicting rates that radionuclides are released from waste packages cannot rest upon empirical extrapolations of laboratory leach data. Reliable predictions can be based on simple bounding theoretical models, such as solubility-limited bulk-flow, if the assumed parameters are reliably known or defensibly conservative. Wherever possible, performance analysis should proceed beyond simple bounding calculations to obtain more realistic - and usually more favorable - estimates of expected performance. Desire for greater realism must be balanced against increasing uncertainties in prediction and loss of reliability. Theoretical predictions of release rate based on mass-transfer analysis are bounding and the theory can be verified. Postulated repository analogues to simulate laboratory leach experiments introduce arbitrary and fictitious repository parameters and are shown not to agree with well-established theory. 34 refs., 3 figs., 2 tabs

  1. Accelerator assisted repositories, technical issue identification and economic assessment

    Energy Technology Data Exchange (ETDEWEB)

    Myers, T.; Favale, A.; Berwald, D. [Northrop Grumman Corp., Bethpage, NY (United States)

    1995-12-31

    An investigation (technical and economic) is presented to quantify the impact of accelerator assisted repository (AAR) system on the US national nuclear repository strategy. An overview of US spent fuel policy is presented which provides a means to comparatively assess competing spent fuel disposition technologies under investigation. The advantages of an AAR system are highlighted and discussed. A companion economic analysis is presented. The results indicate that although a US geologic waste repository will continue to be required, waste partitioning and accelerator transmutation of selected actinides, and long-lived fission products, can result in substantial benefits. (author) 18 refs.

  2. 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)

  3. Assessing the demographic and public service impacts of repository siting

    International Nuclear Information System (INIS)

    Murdock, S.H.; Hamm, R.R.

    1983-01-01

    Demographic and public service impacts are likely to be among the most evident of those changes resulting from nuclear waste repository development. Knowledge of the characteristics of such impacts and of the means to assess them is critical. The first section of this chapter examines those likely to be unique to repositories. The second section describes the alternatives for assessing such impacts and the particular difficulties likely to affect the assessments. Given the state of development of techniques for assessing impacts and the range of factors that must be considered, perhaps their best use is as a means of sensitizing decision makers to the potential implications of their decisions. 2 tables

  4. A GoldSim Based Biosphere Assessment Model for a HLW Repository

    International Nuclear Information System (INIS)

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

    2007-01-01

    To demonstrate the performance of a repository, the dose exposure to a human being due to nuclide releases from a repository should be evaluated and the results compared to the dose limit presented by the regulatory bodies. To evaluate a dose rate to an individual due to a long-term release of nuclides from a HLW repository, biosphere assessment models and their implemented codes such as ACBIO1 and ACBIO2 have been developed with the aid of AMBER during the last few years. BIOMASS methodology has been adopted for a HLW repository currently being considered in Korea, which has a similar concept to the Swedish KBS-3 HLW repository. Recently, not just only for verifying the purpose for biosphere assessment models but also for varying the possible alternatives to assess the consequences in a biosphere due to a HLW repository, another version of the assessment modesl has been newly developed in the frame of development programs for a total system performance assessment modeling tool by utilizing GoldSim. Through a current study, GoldSim approach for a biosphere modeling is introduced. Unlike AMBER by which a compartment scheme can be rather simply constructed with an appropriate transition rate between compartments, GoldSim was designed to facilitate the object-oriented modules by which specific models can be addressed in an additional manner, like solving jig saw puzzles

  5. Consequences of the Thermal Transient on the Evolution of the Damaged Zone Around a Repository for Heat-Emitting High-Level Radioactive Waste in a Clay Formation: a Performance Assessment Perspective

    Science.gov (United States)

    Yu, Li; Weetjens, Eef; Sillen, Xavier; Vietor, Tim; Li, Xiangling; Delage, Pierre; Labiouse, Vincent; Charlier, Robert

    2014-01-01

    A proper evaluation of the perturbations of the host rock induced by the excavation and the emplacement of exothermic wastes is essential for the assessment of the long-term safety of high-level radioactive waste disposals in clay formations. The impact of the thermal transient on the evolution of the damaged zone (DZ) has been explored in the European Commission project TIMODAZ (thermal impact on the damaged zone around a radioactive waste disposal in clay host rocks, 2006-2010). This paper integrates the scientific results of the TIMODAZ project from a performance assessment (PA) point of view, showing how these results support and justify key PA assumptions and the values of PA model parameters. This paper also contextualises the significance of the thermal impact on the DZ from a safety case perspective, highlighting how the project outcomes result into an improved understanding of the thermo-hydro-mechanical behaviour of the clay host rocks. The results obtained in the TIMODAZ project strengthen the assessment basis of the safety evaluation of the current repository designs. There was no evidence throughout the TIMODAZ experimental observations of a temperature-induced additional opening of fractures nor of a significant permeability increase of the DZ. Instead, thermally induced plasticity, swelling and creep seem to be beneficial to the sealing of fractures and to the recovery of a very low permeability in the DZ, close to that of an undisturbed clay host rock. Results from the TIMODAZ project indicate that the favourable properties of the clay host rock, which guarantee the effectiveness of the safety functions of the repository system, are expected to be maintained after the heating-cooling cycle. Hence, the basic assumptions usually made in PA calculations so far are expected to remain valid, and the performance of the system should not be affected in a negative way by the thermal evolution of the DZ around a radioactive waste repository in clay host rock.

  6. 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

  7. 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.

  8. Simultaneous resolution of reactive radioactive decay, non-isothermal flow, and migration with application to the performance assessment for HLW repositories

    International Nuclear Information System (INIS)

    Juncosa, R.; Delgado, J.; Font, I.

    2010-01-01

    Radioactive decay is an important subject to take into account when studying the thermo-hydro-dynamic behavior of the buffer clay material used in the containment of radioactive waste. The modern concepts for the multibarrier design of a repository of high level waste in deep geologic formations consider that once canisters have failed, the buffer clay material must ensure the retention and/or delay of radionuclides within the time framework given in the assessment studies. Within the clay buffer, different chemical species are retarded/fixed according to several physicochemical processes (ion exchange, surface complexation, precipitation, matrix diffusion,..) but typical approaches do not consider the eventuality that radioactive species change their chemical nature (i.e. phase) thereby affecting their reactive behavior. The radioactive decay of an element takes place independently of the phase (aqueous, solid or gaseous) to which it belongs. This means that, in terms of radionuclide fixation, some geochemical processes will be effective scavengers (for instance mineral precipitation of crystal growth) while others will not (for instance ion exchange and/or sorption). In this contribution we present a reactive radioactive decay model of any number of chemical components including those that belong to decay series. The model, which is named FLOW-DECAY, also takes into account flow and isotopic migration and it has been applied considering a hypothetical model scenario provided by the project ENRESA 2000 and direct comparison with the results generated by the probabilistic code GoldSim. Results indicate that FLOW-DECAY may simulate the decay processes in a similar way that GoldSim, being the differences related to factors associated to code architecture. (orig.)

  9. Parametric study of geohydrologic performance characteristics for geologic waste repositories

    International Nuclear Information System (INIS)

    Bailey, C.E.; Marine, I.W.

    1980-11-01

    One of the major objectives of the National Waste Terminal Storage Program is to identify potential geologic sites for storage and isolation of radioactive waste (and possibly irradiated fuel). Potential sites for the storage and isolation of radioactive waste or spent fuel in a geologic rock unit are being carefully evaluated to ensure that radionuclides from the stored waste or fuel will never appear in the biosphere in amounts that would constitute a hazard to the health and safety of the public. The objective of this report is to quantify and present in graphical form the effects of significant geohydrologic and other performance characteristics that would influence the movement of radionuclides from a storage site in a rock unit to the biosphere. The effort in this study was focused on transport by groundwater because that is the most likely method of radionuclide escape. Graphs of the major performance characteristics that influence the transport of radionuclides from a repository to the biosphere by groundwater are presented. The major characteristics addressed are radioactive decay, leach rate, hydraulic conductivity, porosity, groundwater gradient, hydrodynamic dispersion, ion exchange, and distance to the biosphere. These major performance characteristics are combind with each other and with the results of certain other combinations and presented in graphical form to provide the interrelationships of values measured during field studies. The graphical form of presentation should be useful in the screening process of site selection. An appendix illustrates the use of these graphs to assess the suitability of a site

  10. 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

  11. Assessment of the Durability of Cementitious Materials in Repository Environment

    International Nuclear Information System (INIS)

    Vicente, R.; Marumo, J.T.; Miyamoto, H.; Isiki, V.L.K.; Ferreira, E.G.

    2013-01-01

    The Radioactive Waste Management Laboratory of the Energy and Nuclear Research Institute is developing the concept of a borehole repository for disused sealed radioactive sources drilled in a deep granite batholite. In this concept, the annular space between the well steel casing and the geological formation is backfilled with cement paste. The hardened cement paste functions as an additional barrier against the escape of radionuclides from the repository and their migration to the environment. It also functions as an obstacle to the flow of groundwater between different layers of the geological setting crossed by the borehole. The long term behavior of hydrated cement compounds is yet incompletely known and therefore more research is needed to increase the confidence on the performance of the material under the repository conditions as required. For the repository to achieve the required performance, the cement paste must be durable. However, in a deep repository, the cementitious materials is exposed to the deleterious action of high temperatures and pressures, the radiation field created by the radioactive sources and aggressive ion species that may be present in groundwater. Furthermore, it is necessary to consider that the cement paste is unstable in the long term because its microstructure and mineralogy change with time as the cement gel components recrystallize and react chemically with materials of the repository environment. In principle, the lifetime of this material could be determined based on the study of its long-term behavior, which, in turn, could be estimated from the extrapolation of short-term results, by accelerating, under controlled laboratory conditions, the composition changes and the loss of mechanical strength and cohesion induced by any detrimental component of the repository environment. Loss of mechanical strength, dimensional variations, changes in chemical-mineralogical composition, and leaching of hydrate compounds are all possible

  12. Advances in the study of far-field phenomena affecting repository performance

    International Nuclear Information System (INIS)

    Tsang, C.F.

    1991-01-01

    Studies of far-field phenomena affecting repository performance have focussed on the role of fractures and other heterogeneities in the potential transport of radioactive solutes from the repository to the biosphere. The present paper summarizes two recent advances in the subject: the channeling model for the understanding and analysis tracer transport in variable-aperture fractures and the modeling of coupled thermo-hydro-mechanical processes in geologic formation around a repository. The paper concludes with remarks on the need for duality in the approach to performance assessment. One line of the duality is fundamental studies and the other, goal-oriented assessment to satisfy regulatory requirements. 15 refs., 10 figs., 1 tab

  13. Methods of calculating the post-closure performance of high-level waste repositories

    International Nuclear Information System (INIS)

    Ross, B.

    1989-02-01

    This report is intended as an overview of post-closure performance assessment methods for high-level radioactive waste repositories and is designed to give the reader a broad sense of the state of the art of this technology. As described here, ''the state of the art'' includes only what has been reported in report, journal, and conference proceedings literature through August 1987. There is a very large literature on the performance of high-level waste repositories. In order to make a review of this breadth manageable, its scope must be carefully defined. The essential principle followed is that only methods of calculating the long-term performance of waste repositories are described. The report is organized to reflect, in a generalized way, the logical order to steps that would be taken in a typical performance assessment. Chapter 2 describes ways of identifying scenarios and estimating their probabilities. Chapter 3 presents models used to determine the physical and chemical environment of a repository, including models of heat transfer, radiation, geochemistry, rock mechanics, brine migration, radiation effects on chemistry, and coupled processes. The next two chapters address the performance of specific barriers to release of radioactivity. Chapter 4 treats engineered barriers, including containers, waste forms, backfills around waste packages, shaft and borehole seals, and repository design features. Chapter 5 discusses natural barriers, including ground water systems and stability of salt formations. The final chapters address optics of general applicability to performance assessment models. Methods of sensitivity and uncertainty analysis are described in Chapter 6, and natural analogues of repositories are treated in Chapter 7. 473 refs., 19 figs., 2 tabs

  14. Methods of calculating the post-closure performance of high-level waste repositories

    Energy Technology Data Exchange (ETDEWEB)

    Ross, B. (ed.)

    1989-02-01

    This report is intended as an overview of post-closure performance assessment methods for high-level radioactive waste repositories and is designed to give the reader a broad sense of the state of the art of this technology. As described here, ''the state of the art'' includes only what has been reported in report, journal, and conference proceedings literature through August 1987. There is a very large literature on the performance of high-level waste repositories. In order to make a review of this breadth manageable, its scope must be carefully defined. The essential principle followed is that only methods of calculating the long-term performance of waste repositories are described. The report is organized to reflect, in a generalized way, the logical order to steps that would be taken in a typical performance assessment. Chapter 2 describes ways of identifying scenarios and estimating their probabilities. Chapter 3 presents models used to determine the physical and chemical environment of a repository, including models of heat transfer, radiation, geochemistry, rock mechanics, brine migration, radiation effects on chemistry, and coupled processes. The next two chapters address the performance of specific barriers to release of radioactivity. Chapter 4 treats engineered barriers, including containers, waste forms, backfills around waste packages, shaft and borehole seals, and repository design features. Chapter 5 discusses natural barriers, including ground water systems and stability of salt formations. The final chapters address optics of general applicability to performance assessment models. Methods of sensitivity and uncertainty analysis are described in Chapter 6, and natural analogues of repositories are treated in Chapter 7. 473 refs., 19 figs., 2 tabs.

  15. 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)

  16. Assessing the quality of repositories of volunteered geographical information

    CSIR Research Space (South Africa)

    Cooper, Antony K

    2012-10-01

    Full Text Available the data are used, lack of involvement by users in developing standards, anonymous VGI contributions, bias in VGI, and that not all aspects of data quality can be assessed quantitatively). Their research shows that these repositories have procedures...

  17. Illustration of sampling-based approaches to the calculation of expected dose in performance assessments for the proposed high level radioactive waste repository at Yucca Mountain, Nevada.

    Energy Technology Data Exchange (ETDEWEB)

    Helton, Jon Craig (Arizona State University, Tempe, AZ); Sallaberry, Cedric J. PhD. (.; .)

    2007-04-01

    A deep geologic repository for high level radioactive waste is under development by the U.S. Department of Energy at Yucca Mountain (YM), Nevada. As mandated in the Energy Policy Act of 1992, the U.S. Environmental Protection Agency (EPA) has promulgated public health and safety standards (i.e., 40 CFR Part 197) for the YM repository, and the U.S. Nuclear Regulatory Commission has promulgated licensing standards (i.e., 10 CFR Parts 2, 19, 20, etc.) consistent with 40 CFR Part 197 that the DOE must establish are met in order for the YM repository to be licensed for operation. Important requirements in 40 CFR Part 197 and 10 CFR Parts 2, 19, 20, etc. relate to the determination of expected (i.e., mean) dose to a reasonably maximally exposed individual (RMEI) and the incorporation of uncertainty into this determination. This presentation describes and illustrates how general and typically nonquantitive statements in 40 CFR Part 197 and 10 CFR Parts 2, 19, 20, etc. can be given a formal mathematical structure that facilitates both the calculation of expected dose to the RMEI and the appropriate separation in this calculation of aleatory uncertainty (i.e., randomness in the properties of future occurrences such as igneous and seismic events) and epistemic uncertainty (i.e., lack of knowledge about quantities that are poorly known but assumed to have constant values in the calculation of expected dose to the RMEI).

  18. Draft environmental assessment: reference repository location, Hanford Site, Washington. Nuclear Waste Policy Act (Section 112)

    International Nuclear Information System (INIS)

    1984-12-01

    In February 1983, the US Department of Energy (DOE) identified a reference repository location at the Hanford Site in Washington as one of nine potentially acceptable sites for a mined geologic repository for spent nuclear fuel and high-level radioactive waste. To determine their suitability, the reference repository location at the Hanford Site and the eight other potentially acceptable sites have been evaluated in accordance with the DOE's General Guidelines for the Recommendation of Sites for Nuclear Waste Repositories. These evaluations are reported in this draft environmental assessment (EA), which is being issued for public review and comment. The DOE findings and determinations that are based on these evaluations are preliminary and subject to public review and comment. A final EA will be prepared after considering the comments received on the draft EA. The reference repository location at Hanford is located in the Columbia Plateau, one of five distinct geohydrologic settings that are being considered for the first repository. On the basis of the evaluations reported in this draft EA, the DOE has found that the reference repository location at Hanford is not disqualified under the guidelines. The DOE has also found that it is suitable for site characterization because the evidence does not support a conclusion that the site will not be able to meet each of the qualifying conditions specified in the guidelines. On the basis of these findings, the DOE is proposing to nominate the reference repository location at Hanford as one of five sites suitable for characterization. Furthermore, having performed a comparative evaluation of the five sites proposed for nomination, the DOE has determined that the reference repository location at Hanford is one of three sites preferred for site characterization

  19. A THM stress-strain framework for modelling the performance of argillaceous materials in deep repositories for radioactive waste

    International Nuclear Information System (INIS)

    Laloui, L.; Francois, B.

    2007-01-01

    In the scenarios for deep, geological nuclear-waste repositories, clayey soils will be hydrated, heated, cooled and dried. The numerical modelling of these mechanical processes is a key issue. Performance assessment of deep repositories for heat-generating radioactive waste would benefit from improvements in mechanical stress-strain constitutive modelling of the coupled thermo-hydro-mechanical behaviour. The presented framework allows progress in understanding the most involved phenomena relevant to nuclear-waste repositories and their coupled nature. It could be used both in the design and in the performance assessment of repositories. It may be applied to disposal in clay formations and to hard-rock repositories where artificially compacted clay is to be used as buffer and backfill. Such a constitutive framework may help in understanding some unexplained or controversial behaviours and in defining experimental programmes to answer key questions. (author)

  20. Thermo-mechanical effects from a KBS-3 type repository. Performance of pillars between repository tunnels

    International Nuclear Information System (INIS)

    Hakami, E.; Olofsson, Stig-Olof

    2000-03-01

    The aim of this study has been to investigate how the rock mass, in the near field of a KBS-3 type repository, will be affected by the excavation of tunnels and deposition holes and the thermal load from the deposited waste. The three-dimensional finite difference program FLAC 3D was used to perform numerical simulation of the rock mass behaviour. The rock mass was modelled as a homogeneous and isotropic continuum. The initial area heat intensity of the repository was assumed to be 10 W/m 2 in all models. The results show that in the middle of the pillar between the repository tunnels the temperature reaches a maximum of about 70 deg C after 55 years of deposition. The extent of areas where the rock is predicted to yield depends on the assumed quality of the rock mass and the initial in-situ stress field. The volume of yielded rock reaches a maximum after about 200 years after deposition. For a rock mass with internal friction angle of 45 deg and cohesion of 5 MPa (using a Mohr-Coulomb material model), the extent of yielded rock is limited to about 1.5 m behind the excavation periphery. The largest rock displacements are found in the tunnel floor at the upper part of the deposition holes. Tension and shear failure in the periphery of the excavations is predicted to occur during the rock excavation, with a depth extension depending on the magnitude and orientation of the in-situ stresses, as well as on the rock mass quality. Both the excavation effects and the then-no-mechanical effects are smallest when the major principal stress is oriented parallel with the deposition tunnels. The maximum convergence between tunnel walls was calculated to occur after 200 years and be about 9 mm, in the model assuming a rock mass with 5 MPa cohesion, 45 deg internal friction angle and maximum horizontal stress perpendicular to the tunnel. In this study confining effects from the buffer and backfill material was neglected. The effective stress concept was used in most of the models

  1. Performance Assessment National Review Group

    International Nuclear Information System (INIS)

    Lieberman, J.A.; Davis, S.N.; Harleman, D.R.F.

    1985-02-01

    Performance assessment involves predicting the potential radiological impact of a nuclear waste disposal system, taking into account all of the natural and engineered components of the system. It includes the analysis and evaluation of predicted system and component performance to determine compliance with regulatory performance criteria. In the context of the nuclear waste management program, performance assessment has five major purposes: to assist in the evaluation and selection of repository sites; to guide the research, development, and testing programs; to assist in the evaluation of repository designs; to assist in the evaluation of the design and performance of engineered barriers; and to show regulatory compliance and support repository licensing. Current performance assessment methodologies are still in the developmental stage. Only the simplest of bounding calculations have produced quantitative predictions of radionuclide releases. The methodologies require considerable extension and validation before they can provide answers suitable for project decisions and licensing. 135 refs., 9 figs., 1 tab

  2. Environmental impact assessments and geological repositories: A model process

    International Nuclear Information System (INIS)

    Webster, S.

    2000-01-01

    In a recent study carried out for the European Commission, the scope and application of environmental impact assessment (EIA) legislation and current EIA practice in European Union Member States and applicant countries of Central and Eastern Europe was investigated, specifically in relation to the geological disposal of radioactive waste. This paper reports the study's investigations into a model approach to EIA in the context of geological repositories, including the role of the assessment in the overall decision processes and public involvement. (author)

  3. 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.

  4. 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.

  5. 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.

  6. Spent fuel and high level waste: Chemical durability and performance under simulated repository conditions. Results of a coordinated research project 1998-2004. Part 1: Contributions by participants in the co-ordinated research project on chemical durability and performance assessment under simulated repository conditions

    International Nuclear Information System (INIS)

    2007-07-01

    This publication contains the results of an IAEA Coordinated Research Project (CRP). It provides a basis for understanding the potential interactions of waste form and repository environment, which is necessary for the development of the design and safety case for deep disposal. Types of high level waste matrices investigated include spent fuel, glasses and ceramics. Of particular interest are the experimental results pertaining to ceramic forms such as SYNROC. This publication also outlines important areas for future work, namely, standardized, collaborative experimental protocols for package-release studies, structured development and calibration of predictive models linking the performance of packaged waste and the repository environment, and studies of the long term behaviour of the wastes, including active waste samples. It comprises 15 contributions of the participants on the Coordinated Research Project which are indexed individually.

  7. Analysis of spent fuel performance in a geologic repository

    International Nuclear Information System (INIS)

    Apted, M.J.; Liebetrau, A.M.; Engel, D.W.; Alexander, D.H.

    1986-04-01

    The Analytical REpository Source-Term (AREST) code developed for the US Department of Energy is being used to assess the time-dependent release rate of radionuclides from spent nuclear fuel disposed in geologic repositories. The Waste Package Release (WPR) submodule of AREST calculates the release from individual waste packages containing spent fuel based on site-specific design, solubility, corrosion, sorption, and mass transfer data. Under the open system conditions of a repository, there are two limiting release mechanisms: surface reaction control and transport control. In addition, a separate release case is defined for soluble radionuclides that are inventory limited. Mass transfer equations for each of these processes are incorporated into AREST. Four separate sources are identified in the AREST code based on inventory and release mechanism: UO 2 matrix (transport limited), gap (inventory limited), grain boundary (inventory limited, combined with gap), and cladding (transport limited). The calculated release of nuclides contained in the matrix (> 90% of the entire inventory) is controlled by UO 2 solubility or the solubility of a nuclide-bearing phase, whichever is lower

  8. Environmental assessment: Reference repository location, Hanford site, Washington

    Energy Technology Data Exchange (ETDEWEB)

    none,

    1986-05-01

    In February 1983, the US Department of Energy (DOE) identified a reference repository location at the Hanford Site in Washington as one of the nine potentially acceptable sites for a mined geologic repository for spent nuclear fuel and high-level radioactive waste. The site is in the Columbia Plateau, which is one of five distinct geohydrologic settings considered for the first repository. To determine their suitability, the Hanford Site and the eight other potentially acceptable sites have been evaluated in accordance with the DOE's General Guidelines for the Recommendation of Sites for the Nuclear Waste Repositories. These evaluations were reported in draft environmental assessments (EAs), which were issued for public review and comment. After considering the comments received on the draft EAs, the DOE prepared the final EAs. On the basis of the evaluations reported in this EA, the DOE has found that the Hanford site is not disqualified under the guidelines. The DOE has also found that it is suitable for site characterization because the evidence does not support a conclusion that the site will not be able to meet each of the qualifying conditions specified in the guidelines. On the basis of these findings, the DOE is nominating the Hanford site as one of five sites suitable for characterization.

  9. Environmental assessment: Reference repository location, Hanford site, Washington

    Energy Technology Data Exchange (ETDEWEB)

    none,

    1986-05-01

    In February 1983, the US Department of Energy (DOE) identified a reference repository location at the Hanford Site in Washington as one of the nine potentially acceptable sites for a mined geologic repository for spent nuclear fuel and high-level radioactive waste. The site is in the Columbia Plateau, which is one of five distinct geohydrologic settings considered for the first repository. To determine their suitability, the Hanford site and the eight other potentially acceptable sites have been evaluated in accordance with the DOE's General Guidelines for the Recommendation of Sites for the Nuclear Waste Repositories. These evaluations were reported in draft environmental assessments (EAs), which were issued for public review and comment. After considering the comments received on the draft EAs, the DOE prepared the final EAs. On the basis of the evaluations reported in this EA, the DOE has found that the Hanford site is not disqualified under the guidelines. The DOE has also found that is is suitable for site characterization because the evidence does not support a conclusion that the site will not be able to meet each of the qualifying conditions specified in the guidelines. On the basis of these findings, the DOE is nominating the Hanford site as one of five sites available for characterization.

  10. Environmental assessment: Reference repository location, Hanford site, Washington

    International Nuclear Information System (INIS)

    1986-05-01

    In February 1983, the US Department of Energy (DOE) identified a reference repository location at the Hanford Site in Washington as one of the nine potentially acceptable sites for a mined geologic repository for spent nuclear fuel and high-level radioactive waste. The site is in the Columbia Plateau, which is one of five distinct geohydrologic settings considered for the first repository. To determine their suitability, the Hanford Site and the eight other potentially acceptable sites have been evaluated in accordance with the DOE's General Guidelines for the Recommendation of Sites for the Nuclear Waste Repositories. These evaluations were reported in draft environmental assessments (EAs), which were issued for public review and comment. After considering the comments received on the draft EAs, the DOE prepared the final EAs. On the basis of the evaluations reported in this EA, the DOE has found that the Hanford site is not disqualified under the guidelines. The DOE has also found that it is suitable for site characterization because the evidence does not support a conclusion that the site will not be able to meet each of the qualifying conditions specified in the guidelines. On the basis of these findings, the DOE is nominating the Hanford site as one of five sites suitable for characterization

  11. Environmental assessment: Reference repository location, Hanford site, Washington

    International Nuclear Information System (INIS)

    1986-05-01

    In February 1983, the US Department of Energy (DOE) identified a reference repository location at the Hanford Site in Washington as one of the nine potentially acceptable sites for a mined geologic repository for spent nuclear fuel and high-level radioactive waste. The site is in the Columbia Plateau, which is one of five distinct geohydrologic settings considered for the first repository. To determine their suitability, the Hanford site and the eight other potentially acceptable sites have been evaluated in accordance with the DOE's General Guidelines for the Recommendation of Sites for the Nuclear Waste Repositories. These evaluations were reported in draft environmental assessments (EAs), which were issued for public review and comment. After considering the comments received on the draft EAs, the DOE prepared the final EAs. On the basis of the evaluations reported in this EA, the DOE has found that the Hanford site is not disqualified under the guidelines. The DOE has also found that is is suitable for site characterization because the evidence does not support a conclusion that the site will not be able to meet each of the qualifying conditions specified in the guidelines. On the basis of these findings, the DOE is nominating the Hanford site as one of five sites available for characterization

  12. PASS - Project on Alternative Systems Study. Performance assessment of bentonite clay barrier in three repository concepts: VDH, KBS-3 and VLH

    International Nuclear Information System (INIS)

    Pusch, R.; Boergesson, L.

    1992-12-01

    The three repository concepts VDH, KBS3 and VLH have been investigated with respect to their functions in short- and long-term perspectives. The study shows that while KBS3 does not require development of new techniques for excavation and application of buffers and canisters, such development is needed for VLH and VDH. The various physical processes in the deployment part of VDH are more critical and less understood than those in KBS3 and VLH, but the sealing effect of the plugged 'low-temperature' part is sufficiently good to make the concept qualify as a candidate. VLH has the highest and KBS3 the lowest temperature and the latter has the highest potential for good long-term function. (59 refs.)

  13. 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

  14. Waste package performance assessment

    International Nuclear Information System (INIS)

    Lester, D.H.

    1981-01-01

    This paper describes work undertaken to assess the life-expectancy and post-failure nuclide release behavior of high-level and waste packages in a geologic repository. The work involved integrating models of individual phenomena (such as heat transfer, corrosion, package deformation, and nuclide transport) and using existing data to make estimates of post-emplacement behavior of waste packages. A package performance assessment code was developed to predict time to package failure in a flooded repository and subsequent transport of nuclides out of the leaking package. The model has been used to evaluate preliminary package designs. The results indicate, that within the limitation of model assumptions and data base, packages lasting a few hundreds of years could be developed. Very long lived packages may be possible but more comprehensive data are needed to confirm this

  15. 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

  16. Melodie: A global risk assessment model for radioactive waste repositories

    International Nuclear Information System (INIS)

    Lewi, J.; Assouline, M.; Bareau, J.; Raimbault, P.

    1987-03-01

    The Institute of Protection and Nuclear Safety (IPSN), which is part of the French Atomic Energy Commission (C.E.A.) develops since 1984 in collaboration with different groups inside and outside the C.E.A. a computer model for risk assessment of nuclear waste repositories in deep geological formations. The main characteristics of the submodels, the data processing structure and some examples of applications are presented

  17. Enhancement of the methodology of repository design and post-closure performance assessment for preliminary investigation stage. Progress report on NUMO-JAEA collaborative research in FY2011 (Joint research)

    International Nuclear Information System (INIS)

    Shibata, Masahiro; Sawada, Atsushi; Tachi, Yukio; Makino, Hitoshi; Hayano, Akira; Mitsui, Seiichiro; Taniguchi, Naoki; Oda, Chie; Kitamura, Akira; Osawa, Hideaki; Semba, Takeshi; Hioki, Kazumasa; Kamei, Gento; Ebashi, Takeshi; Kubota, Shigeru; Kurosawa, Susumu; Goto, Junichi; Goto, Takahiro; Ishii, Eiichi; Inagaki, Manabu; Moriya, Toshifumi; Suzuki, Satoru; Ohi, Takao; Ichihara, Takayuki; Ishida, Keisuke; Ishiguro, Katsuhiko; Tsuchi, Hiroyuki

    2012-09-01

    JAEA and NUMO have conducted a collaborative research work which is designed to enhance the methodology of repository design and performance assessment in preliminary investigation stage. The topics of such joint research are (1) study on selection of host rock, (2) study on development of scenario, (3) study on setting nuclide migration parameters, (4) study on ensuring quality of knowledge. With regard to (1), in terms of hydraulic properties, items for assessing rock property, and assessment methodology of groundwater travel time has been organized with interaction from site investigation. With regard to (2), the existing approach has been embodied, in addition, the phenomenological understanding regarding dissolution of and nuclide release from vitrified waste, corrosion of the overpack, long-term performance of the buffer are summarized. With regard to (3), the approach for parameter setting has been improved for sorption and diffusion coefficient of buffer/rock, and applied and tested for parameter setting of key radionuclides. With regard to (4), framework for ensuring quality of knowledge has been studied and examined aimed at the likely disposal facility condition. (author)

  18. Fundamentals of the NEA Thermochemical Database and its influence over national nuclear programs on the performance assessment of deep geological repositories.

    Science.gov (United States)

    Ragoussi, Maria-Eleni; Costa, Davide

    2017-03-14

    For the last 30 years, the NEA Thermochemical Database (TDB) Project (www.oecd-nea.org/dbtdb/) has been developing a chemical thermodynamic database for elements relevant to the safety of radioactive waste repositories, providing data that are vital to support the geochemical modeling of such systems. The recommended data are selected on the basis of strict review procedures and are characterized by their consistency. The results of these efforts are freely available, and have become an international point of reference in the field. As a result, a number of important national initiatives with regard to waste management programs have used the NEA TDB as their basis, both in terms of recommended data and guidelines. In this article we describe the fundamentals and achievements of the project together with the characteristics of some databases developed in national nuclear waste disposal programs that have been influenced by the NEA TDB. We also give some insights on how this work could be seen as an approach to be used in broader areas of environmental interest. Copyright © 2017 Elsevier Ltd. All rights reserved.

  19. Types of safety assessments of near surface repository for radioactive waste

    International Nuclear Information System (INIS)

    Mateeva, M.

    2004-01-01

    The purpose of this article is to presents the classification of different types safety assessments of near surface repository for low and intermediate level radioactive waste substantiated with results of safety assessments generated in Bulgaria. The different approach of safety assessments applied for old existing repository as well as for site selection for construction new repository is outlined. The regulatory requirements in Bulgaria define three main types of assessments: Safety assessment; Technical substation of repository safety; Assessment of repository influence on environment that is in form of report prepared from the Ministry of environment and waters on the base of results obtained in two first types of assessments. Additionally first type is subdivided in three categories - preliminary safety assessment, safety assessment and post closure safety assessment, which are generated using deterministic approach. The technical substation of repository safety is generated using probabilistic approach. Safety assessment results that are presented here are based on evaluation of existing old repository type 'Radon' in Novi Han and real site selection procedure for new near surface repository for low and intermediate level radioactive waste from nuclear power station in Kozloduy. The important role of safety assessment for improvement the repository safety as well as for repository licensing, correct site selection and right choice of engineer barriers and repository design is discussed using generated results. (author)

  20. Environmental impact assessments and geological repositories for radioactive waste

    International Nuclear Information System (INIS)

    O'Sullivan, P.; McKirdy, B.; Askarieh, M.; Bond, A.; Russell, S.

    1999-01-01

    Since 1985 it has been obligatory that facilities in the European Union designed for the permanent storage or disposal of radioactive waste be assessed to determine their effects on the environment. This assessment must be undertaken in advance of any decision by national authorities to give consent for development work to proceed. Member States are given wide discretion on how the above requirements are implemented in practice, e.g. the relevant European Council Directives call for the results of the environmental assessment to be made available to the public before development consent is granted but the detailed arrangements for dissemination of such information and procedures for public consultation are determined by individual Member States. Although the Directives require an assessment of the direct and indirect effects of a project on human beings and on various elements of the natural environment, they are non-specific as to what particular impacts should be addressed, particularly as regards the effects of a project on human beings. Therefore, for example, each Member State may decide whether or not social, health and economic impacts should be included in the assessment. This paper discusses the above issues. It proposes a model approach to environmental impact assessment in the context of geological repositories, including the role of the assessment on the overall decision processes for repository development, the scope and content of the assessment report, and approaches to public involvement

  1. 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

  2. Cementitious near-field sorption data bases for performance assessment of a L/ILW repository in a Palfris marl host rock. CEM-94: update I, June 1997

    International Nuclear Information System (INIS)

    Bradbury, M.H.; Loon, L.R. van

    1998-01-01

    This report is an update on an earlier cementitious sorption data base (SDB) prepared by Bradbury and Sarott (1994). The aim is to review any new information or data which have become available in the intervening time and modify the existing SDB appropriately. Discussions will be confined predominantly to areas which have led to significant changes to or reappraisals of the data/values or procedures for obtaining/modifying them. From this point of view this update and the previous SDB are closely related and belong together. The complexation of radionuclides with organic ligands from the chemical degradation of cellulose, and the subsequent negative effects on sorption properties, were identified as being processes of great importance. Since 1994 significant progress has been made in this field and a major part of this work is devoted to a reassessment of the impact of 'organics' on near-field sorption. In particular, the very conservative assumptions which had been made previously because of the general lack of good quality data available at that time, could be replaced by realistic parameter estimates based on new knowledge. For example, maximum likely concentrations of cellulose degradation products and cement additives in the cement pore waters could be calculated allowing the potential effects of these organic ligands on sorption to be bounded. Sorption values for safety relevant radionuclides corresponding to the three broad stages of cement/concrete degradation during the lifetime of the repository are presented in tabulated form. The influence of the wide variety of organic ligands existing in the different waste categories, SMA-1 to SMA-4, is quantified in terms of sorption reduction factors. In the compilation of this cement SDB update, radionuclide uptake onto the vast quantities of aggregate materials and corrosion products from iron/steel was not taken into account. (author) 10 figs., 8 tabs., refs

  3. Assessment of Savannah River borosilicate glass in the repository environment

    International Nuclear Information System (INIS)

    Plodinec, M.J.; Wicks, G.G.; Bibler, N.E.

    1982-04-01

    Since 1973, borosilicate glass has been studied as a matrix for the immobilization of high-level radioactive waste generated at the Savannah River Plant (SRP). In 1977, efforts began to develop and test the large-scale equipment necessary to convert the alkaline waste slurries at SRP into a durable borosilicate glass. A process has now been developed for the proposed Defense Waste Processing Facility (DWPF) which will annually produce approximately 500 canisters of SRP waste glass which will be stored on an interim basis on the Savannah River site. Current national policy calls for the permanent disposal of high-level waste in deep geologic repositories. In the repository environment, SRP waste glass will eventually be exposed to such stresses as lithostatic or hydrostatic pressures, radiation fields, and self-heating due to radioactive decay. In addition, producing and handling each canister of glass will also expose the glass to thermal and mechanical stresses. An important objective of the extensive glass characterization and testing programs of the Savannah River Laboratory (SRL) has been to determine how these stresses affect the performance of SRP waste glass. The results of these programs indicate that: these stresses will not significantly affect the performance of borosilicate glass containing SRP waste; and SRP waste glass will effectively immobilize hazardous radionuclides in the repository environment

  4. 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.)

  5. DECOVALEX III/BENCHPAR PROJECTS. Approaches to Upscaling Thermal-Hydro-Mechanical Processes in a Fractured Rock. Mass and its Significance for Large-Scale Repository Performance Assessment. Summary of Findings. Report of BMT2/WP3

    International Nuclear Information System (INIS)

    Andersson, Johan; Staub, Isabelle; Knight, Les

    2005-02-01

    The Benchmark Test 2 of DECOVALEX III and Work Package 3 of BENCHPAR concerns the upscaling Thermal (T), Hydrological (H) and Mechanical (M) processes in a fractured rock mass and its significance for large-scale repository performance assessment. The work is primarily concerned with the extent to which various thermo-hydro-mechanical couplings in a fractured rock mass adjacent to a repository are significant in terms of solute transport typically calculated in large-scale repository performance assessments. Since the presence of even quite small fractures may control the hydraulic, mechanical and coupled hydromechanical behaviour of the rock mass, a key of the work has been to explore the extent to which these can be upscaled and represented by 'equivalent' continuum properties appropriate PA calculations. From these general aims the BMT was set-up as a numerical study of a large scale reference problem. Analysing this reference problem should: help explore how different means of simplifying the geometrical detail of a site, with its implications on model parameters, ('upscaling') impacts model predictions of relevance to repository performance, explore to what extent the THM-coupling needs to be considered in relation to PA-measures, compare the uncertainties in upscaling (both to uncertainty on how to upscale or uncertainty that arises due to the upscaling processes) and consideration of THM couplings with the inherent uncertainty and spatial variability of the site specific data. Furthermore, it has been an essential component of the work that individual teams not only produce numerical results but are forced to make their own judgements and to provide the proper justification for their conclusions based on their analysis. It should also be understood that conclusions drawn will partly be specific to the problem analysed, in particular as it mainly concerns a 2D application. This means that specific conclusions may have limited applicability to real problems in

  6. DECOVALEX III III/BENCHPAR PROJECTS. Approaches to Upscaling Thermal-Hydro-Mechanical Processes in a Fractured Rock. Mass and its Significance for Large-Scale Repository Performance Assessment. Summary of Findings. Report of BMT2/WP3

    Energy Technology Data Exchange (ETDEWEB)

    Andersson, Johan (comp.) [JA Streamflow AB, Aelvsjoe (Sweden); Staub, Isabelle (comp.) [Golder Associates AB, Stockholm (Sweden); Knight, Les (comp.) [Nirex UK Ltd, Oxon (United Kingdom)

    2005-02-15

    The Benchmark Test 2 of DECOVALEX III and Work Package 3 of BENCHPAR concerns the upscaling Thermal (T), Hydrological (H) and Mechanical (M) processes in a fractured rock mass and its significance for large-scale repository performance assessment. The work is primarily concerned with the extent to which various thermo-hydro-mechanical couplings in a fractured rock mass adjacent to a repository are significant in terms of solute transport typically calculated in large-scale repository performance assessments. Since the presence of even quite small fractures may control the hydraulic, mechanical and coupled hydromechanical behaviour of the rock mass, a key of the work has been to explore the extent to which these can be upscaled and represented by 'equivalent' continuum properties appropriate PA calculations. From these general aims the BMT was set-up as a numerical study of a large scale reference problem. Analysing this reference problem should: help explore how different means of simplifying the geometrical detail of a site, with its implications on model parameters, ('upscaling') impacts model predictions of relevance to repository performance, explore to what extent the THM-coupling needs to be considered in relation to PA-measures, compare the uncertainties in upscaling (both to uncertainty on how to upscale or uncertainty that arises due to the upscaling processes) and consideration of THM couplings with the inherent uncertainty and spatial variability of the site specific data. Furthermore, it has been an essential component of the work that individual teams not only produce numerical results but are forced to make their own judgements and to provide the proper justification for their conclusions based on their analysis. It should also be understood that conclusions drawn will partly be specific to the problem analysed, in particular as it mainly concerns a 2D application. This means that specific conclusions may have limited applicability

  7. Total system performance predictions (TSPA-1995) for the potential high-level waste repository at Yucca Mountain

    International Nuclear Information System (INIS)

    Sevougian, S.D.; Andrews, R.W.; McNeish, J.A.

    1996-01-01

    The management and operating contractor for the potential high-level nuclear waste repository at Yucca Mountain, Nevada, has been recently completed a new performance assessment of the ability of the repository to isolate and contain nuclear waste for long time periods (up to 1,000,000 years). Sensitivity analyses determine the most important physical parameters and processes, using the most current information and models

  8. 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

  9. 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.

  10. TURVA-2012: Performance assessment

    International Nuclear Information System (INIS)

    Hellae, Pirjo; Snellman, Margit; Marcos, Nuria; Pastina, Barbara; Smith, Paul; Koskinen, Kari

    2014-01-01

    TURVA-2012 is Posiva's safety case in support of the Preliminary Safety Analysis Report (PSAR) and application for a construction licence for a repository for disposal of spent nuclear fuel at the Olkiluoto site in south-western Finland. Posiva's safety concept is based on long-term isolation and containment, which is achieved through a robust engineered barrier system (EBS) design and favourable geological conditions at the repository site. The reference design considered in the TURVA-2012 safety case is the KBS-3V design, with the EBS consisting of a copper-iron canister, a buffer of swelling clay material, a backfill in the deposition tunnels of low-permeability material and closure of the central tunnels and other underground openings. The host rock acts as a natural barrier. Each barrier contributes to safety through one or more safety function. The conditions needed for the barriers to fulfil their respective safety functions are expressed in terms of performance targets for the EBS and the target properties for the host rock. The performance assessment (Posiva, 2013), which is a key component of TURVA-2012, analyses the ability of the repository system to provide containment and isolation of the spent nuclear fuel during the long-term evolution of the system and the site. The conditions needed for the barriers to fulfil their respective safety functions are expressed in terms of performance targets for the engineered barriers and target properties for the host rock, for example properties related to the corrosion resistance and mechanical strength of the canister as well as groundwater flow and composition. The analyses take into account the uncertainties in the initial state, the subsequent thermal, hydraulic, mechanical and chemical evolution of the repository system and uncertainties in the evolution. The conclusions of the performance assessment are based mostly on the output of key modelling activities. Whenever modelling is not possible the conclusions

  11. 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)

  12. Design and Performance Assessment of a Conceptual Cover Cap of Near Surface Repository for Short Lived Low and Intermediate Level Radioactive Waste

    Energy Technology Data Exchange (ETDEWEB)

    Asaba, Ruth; Kim, Changlak [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2014-05-15

    The safety of the overall waste disposal system is determined by the performance of its individual components: waste form, waste container, engineered barrier and the host environment. The design of the cover cap helps in preventing percolation of water, and in retarding radionuclide migration from the disposal facility to the biosphere. The cover of a disposal facility is usually a combination of materials such as sand, gravel, concrete, clay and soil conditioned for vegetation growth. The cover system will be designed using models such as EPA's Hydrological Evaluation of Land fill Performance (HELP) code. This paper describes a conceptual design for a cover cap for a land fill as a preferred disposal facility for low and short lived intermediate radioactive waste in Uganda. Majority of the waste is generated from disused sealed and unsealed radioactive sources arising from medical, industrial applications, and research. Radioactive waste management has raised a lot of concern in both developed and developing countries. Each country has a responsibility to come up with a solution to prevent contamination of the environment and humans from radioactive waste. It is important to have thicker soil layers in cap designs so as to support vegetation growth since some activities such as erosion and settlements are expected. Help simulations in this study will assist to demonstrate that it is possible to design a cover cap which can contain radioactive waste packages for hundreds of years provided the proper institutional and performance monitoring schemes are implemented.

  13. Assessment of Public Acceptability in LILW Repository Site Selection Process in Slovenia

    International Nuclear Information System (INIS)

    Zeleznik, N.; Kralj, M.; Polic, M.; Kos, D.

    2006-01-01

    Slovenian national agency for radioactive waste management ARAO has after longer period of preparation activities started with the more direct work on the site selection process for low and intermediate level waste (LILW) repository. In November 2004, the official administrative procedure for the siting of the repository started with the First public conference on spatial planning issues carried out by the Ministry of Environment and Spatial Planning together with ARAO. Just after the conference the Program for the preparation of the Detailed plan of national importance for the LILW repository was accepted by the Ministry. ARAO invited in the beginning of December 2004 all local communities in Slovenia (except 3 of them which have already refused to cooperate) to participate and volunteer a site or area in their local community for further investigation. The invitation for the application of local communities provided clear instructions on how to participate in further determination of potentially suitable sites and under what conditions. By the beginning of April 2005 ARAO finished the bidding process with 8 applications of local communities which decided to participate in the further site selection for LILW repository. Due to the financial and other limitations (human resources, spatial planning procedure, etc.) only in maximum three local communities further characterization could be performed. Therefore prefeasibility study of all volunteer local communities was conducted in which besides technical, environmental and spatial availability also public acceptability should be assessed. For assessment of public acceptability the methodology has been prepared which includes objective parameters of local environment (such as demographic data, economy, infrastructure and social issues in relation to the repository) as well as subjective values (attitudes of individual groups - opinion makers, politicians and all residents - to the sitting and construction of LILW

  14. Enhancement of the methodology of repository design and post-closure performance assessment for preliminary investigation stage (3). Progress report on NUMO-JAEA collaborative research in FY2013 (Joint research)

    International Nuclear Information System (INIS)

    Shibata, Masahiro; Sawada, Atsushi; Tachi, Yukio; Makino, Hitoshi; Wakasugi, Keiichiro; Mitsui, Seiichiro; Kitamura, Akira; Oda, Chie; Ishidera, Takamitsu; Suyama, Tadahiro; Hatanaka, Koichiro; Kamei, Gento; Yoshikawa, Hideki; Senba, Takeshi; Seo, Toshihiro; Kurosawa, Susumu; Goto, Junichi; Shibutani, Sanae; Goto, Takahiro; Kubota, Shigeru; Inagaki, Manabu; Moriya, Toshifumi; Suzuki, Satoru; Ishida, Keisuke; Nishio, Hikaru; Makiuchi, Akie; Fujihara, Hiroshi

    2015-03-01

    JAEA and NUMO have conducted a collaborative research work which is designed to enhance the methodology of repository design and post-closure performance assessment in preliminary investigation stage. With regard to (1) study on host rock suitability in terms of hydrology, based on some examples of developing method of hydro-geological structure model, acquired knowledge are arranged using the tree diagram, and model uncertainty and its influence on the evaluation items were discussed. With regard to (2) study on scenario development, the developed approach for “defining conditions” has been reevaluated and improved from practical viewpoints. In addition, the uncertainty evaluation for the effect of use of cementitious material, as well as glass dissolution model, was conducted with analytical evaluation. With regard to (3) study on setting radionuclide migration parameters, based on survey of precedent procedures, multiple-approach for distribution coefficient of rocks was established, and the adequacy of the approach was confirmed through its application to sedimentary rock and granitic rock. Besides, an approach for solubility setting was developed including the procedure of selection of solubility limiting solid phase. The adequacy of the approach was confirmed through its application to key radionuclides. (author)

  15. 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

  16. 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

  17. 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

  18. 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

  19. Assessment of dose conversion factors in a generic biosphere of a Korea HLW repository

    International Nuclear Information System (INIS)

    Hwang, Y. S.; Park, J. B.; Kang, C. H.

    2002-01-01

    Radioactive species released from a waste repository migrate through engineered and natural barriers and eventually reach the biosphere. Once entered the biosphere, contaminants transport various exposure pathways and finally reach a human. In this study the full RES matrix explaining the key compartments in the biosphere and their interactions is introduced considering the characteristics of the Korean biosphere. Then the three exposure groups are identified based on the compartments of interest. The full exposure pathways and corresponding mathematical expression for mass transfer coefficients and etc are developed and applied to assess the dose conversion factors of nuclides for a specific exposure group. Dose conversion factors assessed in this study will be used for total system performance assessment of a potential Korean HLW repository

  20. Assessing accuracy of an electronic provincial medication repository

    Directory of Open Access Journals (Sweden)

    Price Morgan

    2012-05-01

    Full Text Available Abstract Background Jurisdictional drug information systems are being implemented in many regions around the world. British Columbia, Canada has had a provincial medication dispensing record, PharmaNet, system since 1995. Little is known about how accurately PharmaNet reflects actual medication usage. Methods This prospective, multi-centre study compared pharmacist collected Best Possible Medication Histories (BPMH to PharmaNet profiles to assess accuracy of the PharmaNet profiles for patients receiving a BPMH as part of clinical care. A review panel examined the anonymized BPMHs and discrepancies to estimate clinical significance of discrepancies. Results 16% of medication profiles were accurate, with 48% of the discrepant profiles considered potentially clinically significant by the clinical review panel. Cardiac medications tended to be more accurate (e.g. ramipril was accurate >90% of the time, while insulin, warfarin, salbutamol and pain relief medications were often inaccurate (80–85% of the time. 1215 sequential BPMHs were collected and reviewed for this study. Conclusions The PharmaNet medication repository has a low accuracy and should be used in conjunction with other sources for medication histories for clinical or research purposes. This finding is consistent with other, smaller medication repository accuracy studies in other jurisdictions. Our study highlights specific medications that tend to be lower in accuracy.

  1. Uncertainties in repository performance from spatial variability of hydraulic conductivities - statistical estimation and stochastic simulation using PROPER

    International Nuclear Information System (INIS)

    Lovius, L.; Norman, S.; Kjellbert, N.

    1990-02-01

    An assessment has been made of the impact of spatial variability on the performance of a KBS-3 type repository. The uncertainties in geohydrologically related performance measures have been investigated using conductivity data from one of the Swedish study sites. The analysis was carried out with the PROPER code and the FSCF10 submodel. (authors)

  2. Cryptonite: A Secure and Performant Data Repository on Public Clouds

    Energy Technology Data Exchange (ETDEWEB)

    Kumbhare, Alok; Simmhan, Yogesh; Prasanna, Viktor

    2012-06-29

    Cloud storage has become immensely popular for maintaining synchronized copies of files and for sharing documents with collaborators. However, there is heightened concern about the security and privacy of Cloud-hosted data due to the shared infrastructure model and an implicit trust in the service providers. Emerging needs of secure data storage and sharing for domains like Smart Power Grids, which deal with sensitive consumer data, require the persistence and availability of Cloud storage but with client-controlled security and encryption, low key management overhead, and minimal performance costs. Cryptonite is a secure Cloud storage repository that addresses these requirements using a StrongBox model for shared key management.We describe the Cryptonite service and desktop client, discuss performance optimizations, and provide an empirical analysis of the improvements. Our experiments shows that Cryptonite clients achieve a 40% improvement in file upload bandwidth over plaintext storage using the Azure Storage Client API despite the added security benefits, while our file download performance is 5 times faster than the baseline for files greater than 100MB.

  3. Review on technical issues influencing the performance of chemical barriers of TRU waste repository

    International Nuclear Information System (INIS)

    Fujita, Tomonari; Sugiyama, Daisuke; Tsukamoto, Masaki; Yokoyama, Hayaichi

    1997-01-01

    Studies of technical issues influencing the performance assessment of TRU waste disposal which is occurred from the nuclear fuel reprocessing were reviewed in related to the development of safety analysis method. Especially, the chemical containment was investigated as a key barrier to radionuclide migration. TRU waste including long-lived radionuclides need long-term performance assessment which could be assumed only by the chemical barrier. The description of technical issues concerned with the performance of TRU waste repository has been divided into the following categories: long-term degradation of cementitious materials as engineered barrier for radionuclide migration, effect of colloids, organic macromolecules and organic degradation products on chemical behavior of radionuclides, gas generation by corrosion of metallic wastes, and effects of microbial activity. Preliminary performance assessment indicated that important factors affecting performance of chemical barriers in near-field were the distribution coefficient and the solubility of radionuclides in near-field groundwater. Therefore, it was identified that key issues associated with performance of chemical barrier were evaluation of (a) the long-term change of distribution coefficient of cementitious material through the degradation under repository condition and (b) chemical speciation change of radionuclides such as increase of solubility by the presence of colloidal-size materials. (author)

  4. 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

  5. Performance assessment: a peer review

    International Nuclear Information System (INIS)

    Lieberman, J.A.; Lee, W.W.L.

    1986-01-01

    This paper describes the rationale, membership, operation and major observations of the Performance Assessment National Review Group. The Group was assembled by Weston at the request of the US Department of Energy Office of Civilian Radioactive Waste Management to review performance assessment work in the US basalt, salt and tuff repository projects. The purposes were to evaluate the adequacy of the current methods, identify deficiencies, and suggest potential improvement on repository performance assessment. To perform the review, Weston retained a group of distinguished consultants who have had extensive experience in disciplines pertinent to management of radioactive wastes including mathematical modeling of fluid transport. Topics reviewed included flow and transport, source term and uncertainty analysis. While the emphasis was on methodologies, the Projects were specifically requested to show currently available results so that the way they utilized familiar methodologies could be evaluated. This paper will highlight some of the technical observations of the Group as well as some managerial and institutional issues

  6. Assessment of disruptive scenarios of a Canadian used fuel repository in crystalline rock

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-06-15

    The NWMO has recently extended its modelling capabilities by performing simulations for four disruptive scenarios that, to date, have not yet been examined in detail. These scenarios complement those considered in an existing postclosure safety assessment for a conceptual geological repository located in a hypothetical crystalline rock formation. The four new disruptive scenarios are: Shaft Seal Failure, Undetected Fault, Open or Poorly Sealed Borehole and Open Borehole Due to Inadvertent Human Intrusion. All simulations are based on the FRAC3DVS-OPG Site-Scale Model. The Site-Scale Model includes a simplified representation of the full repository and a portion of the surrounding sub-regional flow system. All transport simulations are performed with only the radionuclide I-129. Transport rates to the surface and a domestic water supply well are compared to the Reference Case results from an earlier case study documented in Reference. (author)

  7. Assessment of disruptive scenarios of a Canadian used fuel repository in crystalline rock

    Energy Technology Data Exchange (ETDEWEB)

    Gobien, M.; Garisto, F.; Hunt, N.; Kremer, E.P., E-mail: mgobien@nwmo.ca [Nuclear Waste Management Organization, Toronto, ON (Canada)

    2015-07-01

    The NWMO has recently extended its modelling capabilities by performing simulations for four disruptive scenarios that, to date, have not yet been examined in detail. These scenarios complement those considered in an existing postclosure safety assessment for a conceptual geological repository located in a hypothetical crystalline rock formation. The four new disruptive scenarios are: Shaft Seal Failure, Undetected Fault, Open or Poorly Sealed Borehole and Open Borehole Due to Inadvertent Human Intrusion. All simulations are based on the FRAC3DVS-OPG [1] Site-Scale Model [2]. The Site-Scale Model includes a simplified representation of the full repository and a portion of the surrounding sub-regional flow system. All transport simulations are performed with only the radionuclide I-129. Transport rates to the surface and a domestic water supply well are compared to the Reference Case results from an earlier case study documented in Reference [2]. (author)

  8. 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)

  9. Assessment of disruptive scenarios of a Canadian used fuel repository in crystalline rock

    International Nuclear Information System (INIS)

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

    2015-01-01

    The NWMO has recently extended its modelling capabilities by performing simulations for four disruptive scenarios that, to date, have not yet been examined in detail. These scenarios complement those considered in an existing postclosure safety assessment for a conceptual geological repository located in a hypothetical crystalline rock formation. The four new disruptive scenarios are: Shaft Seal Failure, Undetected Fault, Open or Poorly Sealed Borehole and Open Borehole Due to Inadvertent Human Intrusion. All simulations are based on the FRAC3DVS-OPG [1] Site-Scale Model [2]. The Site-Scale Model includes a simplified representation of the full repository and a portion of the surrounding sub-regional flow system. All transport simulations are performed with only the radionuclide I-129. Transport rates to the surface and a domestic water supply well are compared to the Reference Case results from an earlier case study documented in Reference [2]. (author)

  10. Approach for Assessing Human Intrusion into a Radwaste Repository

    International Nuclear Information System (INIS)

    Cho, Dong Keun; Kim, Jung Woo; Jeong, Jong Tae; Baik, Min Hoon

    2016-01-01

    An approach to assess human intrusion into radwaste repository resulting from future human actions was proposed based on the common principals, requirements, and recommendations from IAEA, ICRP, and OECD/NEA, with the assumption that the intrusion occurs after loss of knowledge of the hazardous nature of the disposal facility. At first, the essential boundary conditions were derived on the basis of international recommendations, followed by overall approach to deal with inadvertent human intrusion. The essential premises were derived on the basis of international recommendations, followed by overall approach to deal with inadvertent human intrusion. The procedure to derive protective measures was also explained with four steps regarding how to derive safety framework, general measures, potential measures, and eventual protective measures on the basis of stylized scenarios. It is expected that the approach proposed in this study will be effectively used to reduce the potential for and/or consequence of human intrusion during entire processes of realization of disposal facility.

  11. Shear-induced Fracture Slip and Permeability Change. Implications for Long-term Performance of a Deep Geological Repository

    International Nuclear Information System (INIS)

    Min, Ki-Bok; Stephansson, Ove

    2009-03-01

    Opening of fractures induced by shear dilation or normal deformation can be a significant source of fracture permeability change in jointed rock, which is important for the performance assessment of geological repositories for spent nuclear fuel. As the repository generates heat and later cools the fluid-carrying ability of the rocks becomes a dynamic variable during the lifespan of the repository. Heating causes expansion of the rock close to the repository and, at the same time, contraction close to the surface. During the cooling phase of the repository, the opposite takes place. Heating and cooling together with the virgin stress can induce shear dilation of fractures and deformation zones and change the flow field around the repository. The objectives of this project are to examine the contribution of thermal stress to the shear slip of fracture in mid- and far-field around a KBS-3 type of repository and to investigate the effect of evolution of stress on the rock mass permeability. The first part of the study is about the evolution of thermal stresses in the rock during the lifetime of the repository. Critical sections of heat generated stresses around the repository are selected and classified. Fracture data from Forsmark is used to establish fracture network models (DFN) and the models are subjected to the sum of virgin stress and thermal stresses and the shear slip and related permeability change are studied. In the first part of this study, zones of fracture shear slip were examined by conducting a three-dimensional, thermo-mechanical analysis of a spent fuel repository model. Stress evolutions of importance for fracture shear slip are: (1) comparatively high horizontal compressive thermal stress at the repository level, (2) generation of vertical tensile thermal stress right above the repository, (3) horizontal tensile stress near the surface, which can induce tensile failure, and generation of shear stresses at the corners of the repository. In the

  12. Shear-induced Fracture Slip and Permeability Change. Implications for Long-term Performance of a Deep Geological Repository

    Energy Technology Data Exchange (ETDEWEB)

    Min, Ki-Bok (School of Civil, Environmental and Mining Engineering, Univ. of Adelaide, Adelaide (Australia)); Stephansson, Ove (Steph Rock Consulting AB, Berlin (Germany))

    2009-03-15

    Opening of fractures induced by shear dilation or normal deformation can be a significant source of fracture permeability change in jointed rock, which is important for the performance assessment of geological repositories for spent nuclear fuel. As the repository generates heat and later cools the fluid-carrying ability of the rocks becomes a dynamic variable during the lifespan of the repository. Heating causes expansion of the rock close to the repository and, at the same time, contraction close to the surface. During the cooling phase of the repository, the opposite takes place. Heating and cooling together with the virgin stress can induce shear dilation of fractures and deformation zones and change the flow field around the repository. The objectives of this project are to examine the contribution of thermal stress to the shear slip of fracture in mid- and far-field around a KBS-3 type of repository and to investigate the effect of evolution of stress on the rock mass permeability. The first part of the study is about the evolution of thermal stresses in the rock during the lifetime of the repository. Critical sections of heat generated stresses around the repository are selected and classified. Fracture data from Forsmark is used to establish fracture network models (DFN) and the models are subjected to the sum of virgin stress and thermal stresses and the shear slip and related permeability change are studied. In the first part of this study, zones of fracture shear slip were examined by conducting a three-dimensional, thermo-mechanical analysis of a spent fuel repository model. Stress evolutions of importance for fracture shear slip are: (1) comparatively high horizontal compressive thermal stress at the repository level, (2) generation of vertical tensile thermal stress right above the repository, (3) horizontal tensile stress near the surface, which can induce tensile failure, and generation of shear stresses at the corners of the repository. In the

  13. Performance analysis of conceptual waste package designs in salt repositories

    International Nuclear Information System (INIS)

    Jansen, G. Jr.; Raines, G.E.; Kircher, J.F.

    1984-01-01

    A performance analysis of commercial high-level waste and spent fuel conceptual package designs in reference repositories in three salt formations was conducted with the WAPPA waste package code. Expected conditions for temperature, stress, brine composition, radiation level, and brine flow rate were used as boundary conditions to compute expected corrosion of a thick-walled overpack of 1025 wrought steel. In all salt formations corrosion by low Mg salt-dissolution brines typical of intrusion scenarios was too slow to cause the package to fail for thousands of years after burial. In high Mg brines judged typical of thermally migrating brines in bedded salt formations, corrosion rates which would otherwise have caused the packages to fail within a few hundred years were limited by brine availability. All of the brine reaching the package was consumed by reaction with the iron in the overpack, thus preventing further corrosion. Uniform brine distribution over the package surface was an important factor in predicting long package lifetimes for the high Mg brines. 14 references, 15 figures

  14. Assessment of potential perturbations to Posiva's SF repository at Olkiluoto from the ONKALO Facility

    International Nuclear Information System (INIS)

    Alexander, W.R.; Neall, F.B.

    2007-06-01

    . Although outwith the remit of this report, operational safety is also considered due to the potential of accidents and incidents to significantly disrupt the programme. Note that the impacts of constructing the SF repository on the Olkiluoto area are covered elsewhere and neither this nor the repository SA will be addressed here. Rather, the focus of this report is on ONKALO-specific issues that could impact the performance of the Olkiluoto SF repository. Where possible, potentially detrimental phenomena are identified and methods to avoid such issues proposed. Lastly, due to the fact that construction of the ONKALO facility is still ongoing, there remain many uncertainties regarding the impact on the site, so the assessment of any perturbations considered here are generally on the conservative side so that worse-case scenarios for Olkiluoto can be obtained. (orig.)

  15. 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.

  16. Viability Assessment of a Repository at Yucca Mountain. Volume 1: Introduction and Site Characteristics

    Energy Technology Data Exchange (ETDEWEB)

    None

    1998-12-01

    This first volume contains an introduction to the viability assessment, including the purpose, scope, waste forms, technical challenges, an historical perspective, regulatory framework, management of the repository, technical components, preparations for the license application, and repository milestones after the assessment. The second part of this first volume addresses characteristics of the Yucca Mountain site.

  17. A performance goal-based seismic design philosophy for waste repository facilities

    International Nuclear Information System (INIS)

    Hossain, Q.A.

    1994-02-01

    A performance goal-based seismic design philosophy, compatible with DOE's present natural phenomena hazards mitigation and ''graded approach'' philosophy, has been proposed for high level nuclear waste repository facilities. The rationale, evolution, and the desirable features of this method have been described. Why and how the method should and can be applied to the design of a repository facility are also discussed

  18. 10 CFR 60.112 - Overall system performance objective for the geologic repository after permanent closure.

    Science.gov (United States)

    2010-01-01

    ... repository after permanent closure. 60.112 Section 60.112 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) DISPOSAL OF HIGH-LEVEL RADIOACTIVE WASTES IN GEOLOGIC REPOSITORIES Technical Criteria Performance... environment following permanent closure conform to such generally applicable environmental standards for...

  19. Assessment of backfill design for KBS-3V repository

    International Nuclear Information System (INIS)

    Keto, Paula; Dixon, David; Jonsson, Esther; Gunnarsson, David; Boergesson, Lennart; Hansen, Johanna

    2009-12-01

    Posiva and SKB initiated a joint programme BACLO (Backfilling and Closure of the Deep repository) in 2003 with the aim to develop methods and materials for backfilling of deposition tunnels. This report summarises the work done in the third and final phase of the BACLO programme. The main objective of this phase was to study how the various processes active during backfill installation and saturation as well as technical constraints affect its design basis. The work focused on the performance and technical feasibility of a block backfill concept, which calls for filling the majority of the tunnel volume with pre-compacted backfill blocks and the remaining volume with bentonite pellets. Several backfill composition alternatives were chosen for study and they consisted of clay materials with differing amounts of swelling minerals. A large body of information was gained on the effect of different processes on the performance of these backfill options, e.g. water inflow, piping, erosion, self-healing, homogenisation and interaction between backfill and buffer in various laboratory and small-scale field tests. More practical tests included e.g. studies how the blocks and pellets could be installed to the deposition tunnel. Based on the new information on the effect of the processes investigated and the estimated achievable block filling degree and backfill density, recommendations were made concerning material selection, backfill layout and technical issues. In addition, issues requiring further attention to verify the long-term performance of the proposed backfill concept are identified and listed

  20. Assessment of backfill design for KBS-3V repository

    Energy Technology Data Exchange (ETDEWEB)

    Keto, Paula (B+tech, Eurajoki (Finland)); Dixon, David (AECL, Harwell (United Kingdom)); Jonsson, Esther; Gunnarsson, David (Swedish Nuclear Fuel and Waste Management Co., Stockholm (Sweden)); Boergesson, Lennart (Clay Technology AB, Lund (Sweden)); Hansen, Johanna (Posiva (Finland))

    2009-12-15

    Posiva and SKB initiated a joint programme BACLO (Backfilling and Closure of the Deep repository) in 2003 with the aim to develop methods and materials for backfilling of deposition tunnels. This report summarises the work done in the third and final phase of the BACLO programme. The main objective of this phase was to study how the various processes active during backfill installation and saturation as well as technical constraints affect its design basis. The work focused on the performance and technical feasibility of a block backfill concept, which calls for filling the majority of the tunnel volume with pre-compacted backfill blocks and the remaining volume with bentonite pellets. Several backfill composition alternatives were chosen for study and they consisted of clay materials with differing amounts of swelling minerals. A large body of information was gained on the effect of different processes on the performance of these backfill options, e.g. water inflow, piping, erosion, self-healing, homogenisation and interaction between backfill and buffer in various laboratory and small-scale field tests. More practical tests included e.g. studies how the blocks and pellets could be installed to the deposition tunnel. Based on the new information on the effect of the processes investigated and the estimated achievable block filling degree and backfill density, recommendations were made concerning material selection, backfill layout and technical issues. In addition, issues requiring further attention to verify the long-term performance of the proposed backfill concept are identified and listed

  1. MAJOR REPOSITORY DESIGN ISSUES

    International Nuclear Information System (INIS)

    JACK N. BAILEY, DWAYNE CHESTNUT, JAMES COMPTON AND RICHARD D. SNELL

    1997-01-01

    The Yucca Mountain Project is focused on producing a four-part viability assessment in late FY98. Its four components (design, performance assessment, cost estimate, and licensing development plan) must be consistent. As a tool to compare design and performance assessment options, a series of repository pictures were developed for the sequential time phases of a repository. The boundaries of the time phases correspond to evolution in the engineered barrier system (EBS)

  2. Rad waste disposal safety analysis / Integrated safety assessment of a waste repository

    International Nuclear Information System (INIS)

    Jeong, Jongtae; Choi, Jongwon; Kang, Chulhyung

    2012-04-01

    We developed CYPRUS+and adopted PID and RES method for the development of scenario. Safety performance assessment program was developed using GoldSim for the safety assessment of disposal system for the disposal of spnet fuels and wastes resulting from the pyrpoprocessing. Biosphere model was developed and verified in cooperation with JAEA. The capability to evaluate post-closure performance and safety was added to the previously developed program. And, nuclide migration and release to the biosphere considering site characteristics was evaluated by using deterministic and probabilistic approach. Operational safety assessment for drop, fire, and earthquake was also statistically evaluated considering well-established input parameter distribution. Conservative assessment showed that dose rate is below the limit value of low- and intermediate-level repository. Gas generation mechanism within engineered barrier was defined and its influence on safety was evaluated. We made probabilistic safety assessment by obtaining the probability distribution functions of important input variables and also made a sensitivity analysis. The maximum annual dose rate was shown to be below the safety limit value of 10 mSv/yr. The structure and element of safety case was developed to increase reliability of safety assessment methodology for a deep geological repository. Finally, milestone for safety case development and implementation strategy for each safety case element was also proposed

  3. Performance assessment calculational exercises

    International Nuclear Information System (INIS)

    Barnard, R.W.; Dockery, H.A.

    1990-01-01

    The Performance Assessment Calculational Exercises (PACE) are an ongoing effort coordinated by Yucca Mountain Project Office. The objectives of fiscal year 1990 work, termed PACE-90, as outlined in the Department of Energy Performance Assessment (PA) Implementation Plan were to develop PA capabilities among Yucca Mountain Project (YMP) participants by calculating performance of a Yucca Mountain (YM) repository under ''expected'' and also ''disturbed'' conditions, to identify critical elements and processes necessary to assess the performance of YM, and to perform sensitivity studies on key parameters. It was expected that the PACE problems would aid in development of conceptual models and eventual evaluation of site data. The PACE-90 participants calculated transport of a selected set of radionuclides through a portion of Yucca Mountain for a period of 100,000 years. Results include analyses of fluid-flow profiles, development of a source term for radionuclide release, and simulations of contaminant transport in the fluid-flow field. Later work included development of a problem definition for perturbations to the originally modeled conditions and for some parametric sensitivity studies. 3 refs

  4. Life cycle assessment of geological repositories for the final disposal of spent fuel in Finland and Sweden

    International Nuclear Information System (INIS)

    Puhrer, A.; Bauer, C.

    2014-01-01

    This paper presents a Life Cycle Assessment (LCA) of the geological repositories for the final disposal of spent nuclear fuel in Finland and Sweden. A separate LCA has been performed for the geological spent fuel repository in each country and the results have been compared. A further benchmark comparison has been made with the LCA of the Swiss geological repository for high-level waste and spent fuel. The life cycle inventory (LCI) product system boundaries include the spent fuel repository and encapsulation facility in each country. All materials, processes, consumed utilities and transport associated with the construction, operation and closure of the repositories for spent fuel are included in the LCI. The life cycle impact assessment (LCIA) is performed using two methods: IPCC 2007 Climate Change and ReCiPe. These assessment methods return results pertaining to global warming potential (GWP) as well as a number of environmental impact categories such as human toxicity and natural land transformation. Results indicate that the use of copper for disposal canister fabrication and bentonite for repository backfilling are the causes for most of the environmental impact of the spent fuel repositories in Finland and Sweden. Alternate, less bentonite-intensive backfilling scenarios may mitigate this impact. While the Swiss bentonite consumption is lower and no copper is used for canister fabrication, the Swiss electricity and fuel consumption associated with final disposal of high-level waste and spent fuel is significantly higher than in Finland or Sweden. Approximately 1 g CO 2 -eq is emitted due to the final disposal of spent fuel and HLW per kWh of nuclear generated electricity. This represents some 10% of the emissions due to the entire nuclear energy chain and is practically negligible in the context of GHG emissions of other energy technologies. (authors)

  5. Use of One-On Analysis to Evaluate Total System Performance of the Proposed Yucca Mountain Nuclear Waste Repository

    International Nuclear Information System (INIS)

    Saulnier, G.J. Jr.; Lee, K.P.; Mehta, S.; Sevougian, S.D.; Kalinich, D.; McNeish, J.A.

    2002-01-01

    The Yucca Mountain Site Characterization Project is currently evaluating the future performance of the proposed U.S. high-level nuclear waste repository. Using the Total System Performance Assessment (TSPA) model, a stylized analysis was conducted to evaluate the relative importance of natural and engineered barriers to movement of radionuclides from the proposed repository. These stylized ''one-on'' analyses consist of sequentially adding features, components, and processes, associated with the natural and engineered barriers, incorporated within the TSPA model and evaluating the effect of these elements on repository performance, as measured by the total mean annual dose to a reasonably maximally exposed individual. The analyses are ''stylized'' in the sense that they are performed to gain insight only. They are not meant to represent a real physical system in most cases, and in some cases allow the TSPA model to simulate results using parameter ranges outside the normal bounds of the TSPA model. In particular, the analyses provide insight into the relative contributions of repository features and processes in a way that is not possible using the full TSPA performance-assessment model. For example, in the nominal scenario of the TSPA model, the contribution of the natural system is masked by the contribution of the engineered system

  6. Environmental assessment overview, Reference repository location, Hanford site, Washington

    International Nuclear Information System (INIS)

    1986-05-01

    In February 1983, the US Department of Energy (DOE) identified a reference repository location at the Hanford Site in Washington as one of the nine potentially acceptable sites for a mined geologic repository for spent nuclear fuel and high-level radioactive waste. The site is in the Columbia Plateau, which is one of five distinct geohydrologic settings considered for the first repository. To determine their suitability, the Hanford site and the eight other potentially acceptable sites have been evaluated in accordance with the DOE's General Guidelines for the Recommendation of Sites for the Nuclear Waste Repositories. On the basis of the evaluations reported in this EA, the DOE has found that the Hanford site is not disqualified under the guidelines. On the basis of these findings, the DOE is nominating the Hanford site as one of five sites suitable for characterization. 3 figs

  7. Preliminary postclosure risk assessment: Yucca Mountain, Nevada, candidate repository site

    International Nuclear Information System (INIS)

    Eslinger, P.W.; Elwood, D.M.; Freshley, M.D.; Reimus, P.W.; Tanner, J.E.; Doctor, P.G.; Engel, D.W.; Liebetrau, A.M.; Strenge, D.L.; Van Luik, A.E.

    1989-10-01

    A study was conducted by the Pacific Northwest Laboratory for the US Department of Energy, Office of Civilian Radioactive Waste Management, to estimate the postclosure risk, in terms of population health effects, of a proposed high-level nuclear waste repository at Yucca Mountain, Nevada. The risk estimates cover a time span of 1 million years following repository closure. Representative disruptive and intrusive events were selected and evaluated in addition to expected conditions. The estimates were generated assuming spent fuel as the waste form and included all important nuclides from inventory, half-life and dose perspectives. The base case results yield an estimate of 36 health effects over the first million years of repository operation. The doses attributed to the repository corresponds to about 0.1 percent of the doses received from natural background radiation. 16 refs., 1 fig

  8. 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

  9. Nuclear waste repository in basalt: preliminary socioeconomic assessment

    International Nuclear Information System (INIS)

    Cluett, C.; Bolton, P.A.; Malhotra, S.; McStay, J.R.; Slingsby, J.A.

    1984-06-01

    This report was prepared as a part of the continuing site characterization activities for a proposed nuclear waste repository in basalt (NWRB) to be located on the Hanford Site near Richland, Washington. The purpose of this study is to assess the social and economic impacts that could be caused by the construction and operation of the proposed NWRB facility. The specific objectives of this study are to describe historical socioeconomic trends in the study area, to describe current conditions, to project future baseline conditions without the NWRB, to project potential impacts due to the proposed NWRB under two alternative regional development scenarios and assess their significance, and to suggest an overall impact management and mitigation strategy. A closely related objective is to assemble a comprehensive socioeconomic data base that can be easily updated for future analyses. This study examines employment, labor supply, population change, housing, local transportation, revenues, and expenditures for public services. This report documents the marked demographic and economic decline that has occurred in the study area since 1981 and concludes that future baseline growth will resume at a relatively slower pace after further expected declines have been experienced through about 1985. The projected socioeconomic impacts of the NWRB development are assessed under two alternative baseline scenarios and are not expected to be significant in either case. With careful planning and attention to impact mitigation, including public participation and interaction with local and regional planning agencies, potential socioeconomic impacts can be anticipated and managed effectively. Recommendations are made for providing frequent updating of the data base and for improving the analysis of socioeconomic impacts. 68 references, 19 figures, 38 tables

  10. Characteristics and behaviour of C-14, Cl-36, Pu-239 and Tc-99 in the Biosphere in the context of performance assessments of geological repositories for high-level radioactive wastes

    International Nuclear Information System (INIS)

    Aguero, A.; Perez-Sanchez, D.; Trueba, C.; Moraleda, M.

    2006-01-01

    The evaluation of the safety of high-level radioactive wastes repositories is a requisite of the radioactive waste management policy. Criteria for the safety of those facilities [IAEA, 1995] require protection of the population and the environment over a time scale appropriate to the characteristics of the disposed wastes. In order to demonstrate compliance with the relevant principles and regulations, there is the requirement for an analysis and quantification of system behavior including the transport of radionuclides from the facility to the biosphere, their transport in the biosphere and the resulting exposure of humans, including the evaluation of radiation doses from both external and internal exposure. Realistic models to simulate the transport of radionuclides in the environment are essential. The biosphere model needs to represent the transport and accumulation of radionuclides in the different media, including characterization of the pathways through which the contaminants give rise to exposure of humans. Such modeling needs to take account of the specific characteristics of the radionuclides involved, which have different physical, chemical and biological characteristics and behaviors and could require conceptually different models to predict their transport through, and distribution in, the relevant environmental media. Key radionuclides that have been shown to contribute most to the doses estimated in high-level radioactive waste disposal performance assessments were identified in [Watkins and Smith, 1999]. The important factors that determine whether a radionuclide will be a key contributor to dose are: its total inventory, half-life, solubility and adsorption potential, and biosphere flux-to-dose conversion factor. Only a limited number of radionuclides are likely to be important for the groundwater release and transport pathway from the point of view of their contribution to annual individual dose to humans: ''237Np (and its progeny), ''129I, ''99Tc

  11. 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

  12. Assessment of volcanic and tectonic hazards to high level radioactive waste repositories

    International Nuclear Information System (INIS)

    Wallmann, P.C.; Miller, I.; Kossik, R.

    1993-01-01

    Golder Associates Inc. (GAI) has developed a computer program (RIP) for performing probabilistic total system performance assessment and site characterization strategy evaluation which can be applied in an iterative manner to evaluate repository site suitability and to guide characterization activities. The performance assessment model incorporated in RIP has three basic component models: (1) waste package behavior, (2) radionuclide transport pathways, and (3) disruptive events. Classes of disruptive events are specified in RIP by (1) a disruption rate (events/yr.), (2) open-quotes event descriptorsclose quotes which describe event characteristics and magnitude, and (3) the consequences associated with an event. One of the strengths of RIP is its flexibility, which allows it to evaluate different sites and conceptual models. Examples of seismic and volcanic disruptive event models constructed by GAI for Yucca Mountain are presented. Analysis of the results of these models indicates that for the simulated models, neither of these event classes significantly impacts the performance of the proposed repository over a 10,000 year time span

  13. An assessment of the effect on Olkiluoto repository capacity achievable with advanced fuel cycles

    Energy Technology Data Exchange (ETDEWEB)

    Juutilainen, P.; Viitanen, T. [VTT Technical Research Centre of Finland, P.O. Box 1000, FI-02044 VTT (Finland)

    2013-07-01

    Previously a few scenarios have been simulated for transition from thermal to fast reactor fleet in Finland in order to determine how much the transuranic inventory could be reduced with the partitioning and transmutation (P-T) technologies. Those calculations, performed with COSI6 code developed by CEA, are extended in the present study, in which the effect of P-T on the capacity of the planned final disposal repository at Olkiluoto (Finland) is evaluated by taking into account the created fission products and transuranic residuals from the reprocessing operations. The decay heat is assumed to be the most restrictive factor in defining the waste disposal packing density. The repository capacity evaluation of this study is based on the comparison of the decay heats produced by the deposited waste in various scenarios. The reference scenario of this article involves only Light Water Reactors (LWR) in an open fuel cycle. The capacity requirement of the geological repository is estimated in a few closed fuel cycle scenarios, all including actinide transmutation with Fast Reactors (FR). The comparison between the P-T scenarios and reference is based on the decay heat production of the deposited waste. The COSI6 code is used for simulations to provide the repository decay heat curves. Applying the closed fuel cycle would change the disposal concept and schedule, because of which it is not quite straightforward to assess the impact of P-T on the capacity. However, it can be concluded that recycling the transuranic nuclides probably decreases the required volume for the disposal, but thermal dimensioning analysis is needed for more specific conclusions.

  14. Assessment of doses to non-human biota: Review of developments and demonstration assessment for Olkiluoto repository

    International Nuclear Information System (INIS)

    Smith, K.; Robinson, C.

    2006-12-01

    This report provides a summary of work commissioned by Posiva Oy and undertaken by Enviros Consulting Ltd to support the development of a strategy for the assessment of environmental impacts from ionising radiation associated with the Olkiluoto waste repository, Finland, as part of the development of the Posiva Safety Case Portfolio. This project included a review of the development of international policies and standards related to protection of biota from the effects of ionizing radiation and of biota assessment methodologies, paying particular attention to those that have been applied to waste repository performance assessments. On the basis of this review, recommendations were developed on the most appropriate methodology to apply in order to assess the impact of radioactive releases from the planned spent fuel repository in Olkiluoto. A test-case was developed, in collaboration with staff from Posiva and Facilia AB, and an assessment was performed. The results and experience of which were analysed and summarised to develop recommendations for a future strategy. The test case highlighted some significant data gaps related to the assessment of impacts to both generic biota types and to interest species. In particular, concentration ratios for generic carnivorous mammals and migratory species such as moose that may consume food from multiple ecosystems and dose conversion factors for large burrowing (i.e. hibernating) mammals. However, in general terms, the dose rates predicted for all organism types were several orders of magnitude below those at which population effects would be expected to be observed and those at which effects on the individual may be anticipated. There would therefore be scope for simplifying the approach applied, although there would be value in performing a sensitivity analysis to ensure that the simplification is applied appropriately. There would also be value in ensuring consistency of the developing approach for non-human biota with

  15. Assessment of doses to non-human biota: Review of developments and demonstration assessment for Olkiluoto repository

    Energy Technology Data Exchange (ETDEWEB)

    Smith, K. [Carol Robinson Enviros Consulting Ltd, Edinburgh (United Kingdom)

    2006-12-15

    This report provides a summary of work commissioned by Posiva Oy and undertaken by Enviros Consulting Ltd to support the development of a strategy for the assessment of environmental impacts from ionising radiation associated with the Olkiluoto waste repository, Finland, as part of the development of the Posiva Safety Case Portfolio. This project included a review of the development of international policies and standards related to protection of biota from the effects of ionizing radiation and of biota assessment methodologies, paying particular attention to those that have been applied to waste repository performance assessments. On the basis of this review, recommendations were developed on the most appropriate methodology to apply in order to assess the impact of radioactive releases from the planned spent fuel repository in Olkiluoto. A test-case was developed, in collaboration with staff from Posiva and Facilia AB, and an assessment was performed. The results and experience of which were analysed and summarised to develop recommendations for a future strategy. The test case highlighted some significant data gaps related to the assessment of impacts to both generic biota types and to interest species. In particular, concentration ratios for generic carnivorous mammals and migratory species such as moose that may consume food from multiple ecosystems and dose conversion factors for large burrowing (i.e. hibernating) mammals. However, in general terms, the dose rates predicted for all organism types were several orders of magnitude below those at which population effects would be expected to be observed and those at which effects on the individual may be anticipated. There would therefore be scope for simplifying the approach applied, although there would be value in performing a sensitivity analysis to ensure that the simplification is applied appropriately. There would also be value in ensuring consistency of the developing approach for non-human biota with

  16. Microbial Influence on the Performance of Subsurface, Salt-Based Radioactive Waste Repositories. An Evaluation Based on Microbial Ecology, Bioenergetics and Projected Repository Conditions

    International Nuclear Information System (INIS)

    Swanson, J.S.; Reed, D.T.; Cherkouk, A.; Arnold, T.; Meleshyn, A.; Patterson, Russ

    2018-01-01

    For the past several decades, the Nuclear Energy Agency Salt Club has been supporting and overseeing the characterisation of rock salt as a potential host rock for deep geological repositories. This extensive evaluation of deep geological settings is aimed at determining - through a multidisciplinary approach - whether specific sites are suitable for radioactive waste disposal. Studying the microbiology of granite, basalt, tuff, and clay formations in both Europe and the United States has been an important part of this investigation, and much has been learnt about the potential influence of microorganisms on repository performance, as well as about deep subsurface microbiology in general. Some uncertainty remains, however, around the effects of microorganisms on salt-based repository performance. Using available information on the microbial ecology of hyper-saline environments, the bioenergetics of survival under high ionic strength conditions and studies related to repository microbiology, this report summarises the potential role of microorganisms in salt-based radioactive waste repositories

  17. Assessing the potentialities of integrated modelling during early phases of siting and design of a geological repository: the REGIME exercise

    Energy Technology Data Exchange (ETDEWEB)

    Genty, A.; Certes, C.; Serres, C.; Besnus, F. [Institut de Radioprotection et de Surete Nucleaire IRSN, 92 - Fontenay aux Roses (France); Fischer-Appelt, K.; Baltes, B.; Rohlig, J. [Gesellschaft fuer Anlagen- und Reaktorsicherheit mbH (GRS), Koeln (Germany)

    2003-01-01

    This paper presents the safety assessment exercise 'REGIME' (Repository Evaluation performed by GRS and IRSN through a Modelling Exercise) performed jointly by GRS and IRSN. The main objective of the project is to test the ability of integrated modelling to contribute to site selection and repository conception in the context of high-level radioactive waste disposal. The project is divided in two parts. Phase 1 consisted in studying different flow patterns in a given geological context. The selected hydrogeological contexts and three site locations potentially favourable for hosting a repository are described. Phase 2, under progress, aims at evaluating the rote of limitation of releases played by the different components of the disposal system taking into account possible dysfunctions. The main issues to be addressed in phase 2, the modelling outline and the scenarios to be studied are presented. (authors)

  18. Application of systems engineering to determine performance requirements for repository waste packages

    International Nuclear Information System (INIS)

    Aitken, E.A.; Stimmell, G.L.

    1987-01-01

    The waste package for a nuclear waste repository in salt must contribute substantially to the performance objectives defined by the Salt Repository Project (SRP) general requirements document governing disposal of high-level waste. The waste package is one of the engineered barriers providing containment. In establishing the performance requirements for a project focused on design and fabrication of the waste package, the systems engineering methodology has been used to translate the hierarchy requirements for the repository system to specific performance requirements for design and fabrication of the waste package, a subsystem of the repository. This activity is ongoing and requires a methodology that provides traceability and is capable of iteration as baseline requirements are refined or changed. The purpose of this summary is to describe the methodology being used and the way it can be applied to similar activities in the nuclear industry

  19. TURVA-2012: Assessment of radionuclide release scenarios for the repository system

    International Nuclear Information System (INIS)

    Smith, Paul; Poteri, Antti; Nordman, Henrik; Cormenzana, Jose Luis; Snellman, Margit; Marcos, Nuria; Hjerpe, Thomas; Koskinen, Lasse

    2014-01-01

    TURVA-2012 is Posiva's safety case in support of the Preliminary Safety Analysis Report (PSAR) and application for a construction licence for a repository for disposal of spent nuclear fuel at the Olkiluoto site in south-western Finland. This paper gives a summary of the analyses of the radionuclide release scenarios formulated in a companion paper, TURVA-2012: Formulation of Radionuclide Release Scenarios (Marcos, 2014). The scenarios and the analyses take into account major uncertainties in the initial state of the barriers and possible paths for the evolution of the repository system identified in a further paper: TURVA-2012: Performance Assessment (Hellae, 2014). For each scenario, calculation cases are analysed to evaluate compliance of the proposed repository with regulatory requirements on radiological protection, as well as to illustrate the impact of specific uncertainties or combinations of uncertainties on the calculated results. Each case illustrates different possibilities for how the repository might evolve and perform over time, taking into account uncertainties in the models and parameter values used to represent radionuclide release, retention and transport and, for biosphere assessment calculation cases, radiation exposure. The calculation cases each address a single, failed canister, where three possible modes of failure are considered: - The presence of an initial defect in the copper overpack of the canister that penetrates the overpack completely (subsequent corrosion of the insert may then lead to an enlargement of the defect). - Corrosion of the copper overpack, which occurs most rapidly in scenarios in which buffer density is reduced, e.g. by erosion. - Shear movements on fractures intersecting the deposition holes. However, the likelihood and consequences of more than one canister failure occurring during the assessment time fame are also considered, generally based on the findings from the single canister calculations. Quantitative

  20. The assessment of human intrusion into underground repositories for radioactive waste Volume 1: Main report

    International Nuclear Information System (INIS)

    Nancarrow, D.J.; Little, R.H.; Asthon, J.; Staunton, G.M.

    1990-01-01

    This report has been prepared with the primary objective of establishing a methodology for the assessment of human intrusion into deep underground repositories for radioactive wastes. The disposal concepts considered are those studied in the performance assessment studies Pagis and Pacoma, coordinated by the CEC. These comprise four types of host rock, namely: clay, granite, salt and the sub-seabed. Following a review of previous assessments of human intrusion, a list of relevant human activities is derived. This forms the basis for detailed characterization of groundwater abstraction and of exploitation of mineral and other resources. Approaches to assessment of intrusion are reviewed and consideration is given to the estimation of probabilities for specific types of intrusion events. Calculational schemes are derived for specific intrusion events and dosimetric factors are presented. A review is also presented of the capacity for reduction of the risks associated with intrusions. Finally, conclusions from the study are presented

  1. The assessment of human intrusion into underground repositories for radioactive waste Volume 2: Appendices

    International Nuclear Information System (INIS)

    Nancarrow, D.J.; Little, R.H.; Ashton, J.; Staunton, G.M.

    1990-01-01

    This report has been prepared with the primary objective of establishing a methodology for the assessment of human intrusion into deep underground repositories for radioactive wastes. The disposal concepts considered are those studied in the performance assessment studies Pagis and Pacoma, coordinated by the CEC. These comprise four types of host rock, namely: clay, granite, salt and the sub-seabed. Following a review of previous assessments of human intrusion, a list of relevant human activities is derived. This forms the basis for detailed characterization of groundwater abstraction and of exploitation of mineral and other resources. Approaches to assessment of intrusion are reviewed and consideration is given to the estimation of probabilities for specific types of intrusion events. Calculational schemes are derived for specific intrusion events and dosimetric factors are presented. A review is also presented of the capacity for reduction of the risks associated with intrusions. Finally, conclusions from the study are presented

  2. 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.

  3. 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.

  4. 5th Total System Performance Assessment Workshop

    International Nuclear Information System (INIS)

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

    2009-07-01

    Research items on safety assessment of high-level waste repository have been proposed by external invited experts outside KAERI and discussed extensively during the annual 5th performance assessment workshop prepared by safety assessment group in KAERI. This could be useful to set up R and D plans necessary for the next phase of mid- and long-term reaserch area regarding the safety assessment of high-level waste repository. Through the research and the presentation, HLW-related research and development area including such specific research items as current status of HLW safety assessment research, current requirement for the licensing of the repository system, priority on research area, data base building for the safety assessment, source-term modeling as well as safety case, among many others, have been discussed and summarized

  5. A performance goal-based seismic design philosophy for waste repository facilities

    International Nuclear Information System (INIS)

    Hossain, Q.A.

    1994-01-01

    A performance goal-based seismic design philosophy, compatible with DOE's present natural phenomena hazards mitigation and open-quotes graded approachclose quotes philosophy, has been proposed for high level nuclear waste repository facilities. The rationale, evolution, and the desirable features of this method have been described. Why and how the method should and can be applied to the design of a repository facility are also discussed

  6. Assessment of Self-Archiving in Institutional Repositories: Across Disciplines

    Science.gov (United States)

    Xia, Jingfeng

    2007-01-01

    This research examined self-archiving practices by four disciplines in seven institutional repositories. By checking each individual item for its metadata and deposition status, the research found that a disciplinary culture is not obviously presented. Rather, self-archiving is regulated by a liaison system and a mandate policy.

  7. Assessment of Effectiveness of Geologic Isolation Systems: REFERENCE SITE INITIAL ASSESSMENT FOR A SALT DOME REPOSITORY

    Energy Technology Data Exchange (ETDEWEB)

    Harwell, M. A.; Brandstetter, A.; Benson, G. L.; Raymond, J. R.; Brandley, D. J.; Serne, R. J.; Soldat, J. K.; Cole, C. R.; Deutsch, W. J.; Gupta, S. K.; Harwell, C. C.; Napier, B. A.; Reisenauer, A. E.; Prater, L. S.; Simmons, C. S.; Strenge, D. L.; Washburn, J. F.; Zellmer, J. T.

    1982-06-01

    As a methodology demonstration for the Office of Nuclear Waste Isolation (ONWI), the Assessment of Effectiveness of Geologic Isolation Systems (AEGIS) Program conducted an initial reference site analysis of the long-term effectiveness of a salt dome repository. The Hainesville Salt Dome in Texas was chosen to be representative of the Gulf Coast interior salt domes; however, the Hainesville Site has been eliminated as a possible nuclear waste repository site. The data used for this exercise are not adequate for an actual assessment, nor have all the parametric analyses been made that would adequately characterize the response of the geosystem surrounding the repository. Additionally, because this was the first exercise of the complete AEGIS and WASTE Rock Interaction Technology (WRIT) methodology, this report provides the initial opportunity for the methodology, specifically applied to a site, to be reviewed by the community outside the AEGIS. The scenario evaluation, as a part of the methodology demonstration, involved consideration of a large variety of potentially disruptive phenomena, which alone or in concert could lead to a breach in a salt dome repository and to a subsequent transport of the radionuclides to the environment. Without waste- and repository-induced effects, no plausible natural geologic events or processes which would compromise the repository integrity could be envisioned over the one-million-year time frame after closure. Near-field (waste- and repository-induced) effects were excluded from consideration in this analysis, but they can be added in future analyses when that methodology development is more complete. The potential for consequential human intrusion into salt domes within a million-year time frame led to the consideration of a solution mining intrusion scenario. The AEGIS staff developed a specific human intrusion scenario at 100 years and 1000 years post-closure, which is one of a whole suite of possible scenarios. This scenario

  8. Supplemental Performance Analyses for the Potential High-Level Nuclear Waste Repository at Yucca Mountain

    International Nuclear Information System (INIS)

    Sevougian, S. D.; McNeish, J. A.; Coppersmith, K.; Jenni, K. E.; Rickertsen, L. D.; Swift, P. N.; Wilson, M. L.

    2002-01-01

    The U.S. Department of Energy (DOE) is considering the possible recommendation of a site at Yucca Mountain, Nevada, for the potential development of a geologic repository for the disposal of high-level radioactive waste and spent nuclear fuel. To facilitate public review and comment, in May 2001 the DOE released the Yucca Mountain Science and Engineering Report (S and ER) (1), which presents technical information supporting the consideration of the possible site recommendation. The report summarizes the results of more than 20 years of scientific and engineering studies. Based on internal reviews of the S and ER and its key supporting references, the Total System Performance Assessment for the Site Recommendation (TSPA-SR) (2) and the Analysis Model Reports and Process Model Reports cited therein, the DOE has recently identified and performed several types of analyses to supplement the treatment of uncertainty in support of the consideration of a possible site recommendation. The results of these new analyses are summarized in the two-volume report entitled FY01 Supplemental Science and Performance Analysis (SSPA) (3,4). The information in this report is intended to supplement, not supplant, the information contained in the S and ER. The DOE recognizes that important uncertainties will always remain in any assessment of the performance of a potential repository over thousands of years (1). One part of the DOE approach to recognizing and managing these uncertainties is a commitment to continued testing and analysis and to the continued evaluation of the technical basis supporting the possible recommendation of the site, such as the analysis contained in the SSPA. The goals of the work described here are to provide insights into the implications of newly quantified uncertainties, updated science, and evaluations of lower operating temperatures on the performance of a potential Yucca Mountain repository and to increase confidence in the results of the TSPA described

  9. 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

  10. Assessment of impact to non-human biota from a generic waste repository in the UK

    International Nuclear Information System (INIS)

    Smith, K.L.; Robinson, C.A.; Jones, S.R.; Batlle, J.V.; Norris, S.

    2008-01-01

    , has now made assessment in this system possible, it should be noted that many of the concentration ratio values have been based on extrapolation, for example from data for similar organisms or from nuclides that are considered to be biogeochemically similar. This exercise also revealed some drawbacks with use of the ERICA methodology and tool for this type of assessment, in addition to the lack of flexibility (discussed above) to deal with the type of complex time series of data that arise from a repository performance assessment. While it was not an issue for this assessment, the exclusion of some important ecosystems from the ERICA tool may restrict its application inn other circumstances, for example where an existing or future ecosystem may include wetlands or mires. Changes in climate, and the associated ecosystems, are also not easily accommodated. However, it is also questionable whether existing data, particularly on concentration ratios, are sufficiently robust to allow differentiation between different climates. In this study, concentration ratios for the user defined organism - the fox - were derived from an alternative data source. The consistency of these data within the default range and their relevance to a particular climate and ecosystem type would merit further consideration. The robustness of individual concentration ratios is clearly dependent upon the derivation approach and the need for further sensitivity analysis was identified. A study to evaluate the pedigree of these data and the sensitivity of assessments of the long-term releases of long-lived radio-nuclides to these values would be valuable. Such a study has been proposed to take place as part of the BIOPROTA programme (author)(tk)

  11. 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)

  12. Draft Technical Position Subtask 1.1: waste package performance after repository closure. Volume 1

    International Nuclear Information System (INIS)

    Davis, M.S.; Schweitzer, D.G.

    1983-08-01

    This document provides guidance to the DOE on the issues and information necessary for the NRC to evaluate waste package performance after repository closure. Minimal performance objectives of the waste package are required by proposed 10 CFR 60. This Draft Technical Position describes the various options available to the DOE for compliance and discusses advantages and disadvantages of various choices. Examples are discussed dealing with demonstrability, predictability and reasonable assurance. The types of performance are considered. The document summarizes presently identified high priority issues needed to evaluate waste package performance after repository closure. 20 references, 7 tables

  13. Basic repository environment assessment design basis, Cypress Creek Dome Site

    International Nuclear Information System (INIS)

    1988-03-01

    This study examines the engineering factors and costs associated with the construction, operation, and decommissioning of a high-level nuclear waste repository in salt in the Gulf Interior Region at Cypress Creek Cone, Mississippi. The study assumes a repository capacity of 36,000 metric tons of heavy metal (MTHM) of unreprocessed spent fuel and 36,000 MTHM of commercial high-level reprocessing waste, along with 7020 canisters of defense high-level reprocessing waste and associated quantities of remote- and contact-handled transuranic waste (TRU). With the exception of TRU, all the waste forms are placed in 300- to 1000-year-life carbon-steel waste packages in a collocated waste handling and packaging facility (WHPF), which is also described. The construction, operation, and decommissioning of the proposed repository is estimated to cost approximately $4.66 billion. Costs include those for the collocated WHPF, engineering, and contingency, but exclude waste from assembly and shipment to the site and waste package fabrication and shipment to the site. These costs reflect the relatively easy access to the site. Construction would require an estimated 7 years. Engineering factors and costs are not strongly influenced by environmental considerations. 53 refs., 24 figs., 10 tabs

  14. Basic repository environmental assessment design basis, Richton Dome site

    International Nuclear Information System (INIS)

    1988-01-01

    This study examines the engineering factors and costs associated with the construction, operation, and decommissioning of a high-level nuclear waste repository in salt in the Gulf Interior Region at Richton Dome in Perry County, Mississippi. The study assumes a repository capacity of 36,000 metric tons of heavy metal (MTHM) of unreprocessed spent fuel and 36,000 MTHM of commercial high-level reprocessing waste, along with 7,020 canisters of defense high-level reprocessing waste and associated quantities of remote-and contact-handled transuranic waste (TRU). With the exception of TRU, all the waste forms are placed in 300- to 1,000-year-life carbon-steel waste packages in a collocated waste handling and packaging facility (WHPF), which is also described. The construction, operation, and decommissioning of the proposed repository is estimated to cost approximately $4.49 billion. Costs include those for the WHPF, engineering, and contingency, but exclude waste form assembly and shipment to the site and waste package fabrication and shipment to the site. These costs reflect the relative average wage rates of the region and the relatively easy access to the site. Construction would require an estimated 6.25 years. Engineering factors and costs are not strongly influenced by environmental considerations. 52 refs., 24 figs., 20 tabs

  15. Basic repository environmental assessment design basis: Deaf Smith County site

    International Nuclear Information System (INIS)

    1988-03-01

    This study examines the engineering factors and costs associated with the construction, operation, and decommissioning of a high-level nuclear waste repository in salt in the Palo Duro Basin in Deaf Smith County, Texas. The study assumes a repository capacity of 36,000 metric tons of heavy metal (MTHM) of unreprocessed spent fuel and 36,000 MTHM of commercial high-level reprocessing waste, along with 7,020 canisters of defense high-level reprocessing waste and associated quantities or remote- and contact-handled transuranic waste (TRU). With the exception of TRU, all the waste forms are placed in 300- to 1000-year-life carbon-steel waste packages in a collocated waste handling and packaging facility (WHPF), which is also described. The construction, operation, and decommissioning of the proposed repository is estimated to cost approximately $4.64 billion. Costs include those for the collocate WHPF, engineering, and contingency, but exclude waste form assembly and shipment to the site and waste package fabrication and shipment to the site. These costs reflect the relative average wage rates of the region, the relatively easy access to the site, and the relatively weak nature of the salt at this site. Construction would require an estimated 7 to 7.5 years. Engineering factors and costs are not strongly influenced by environmental considerations. 62 refs., 24 figs., 20 tabs

  16. Basic repository environmental assessment design basis, Davis Canyon site

    International Nuclear Information System (INIS)

    1984-01-01

    This study examines the engineering factors and costs associated with the construction, operation, and decommissioning of a high-level nuclear waste repository in salt in the Paradox Basin in Davis Canyon, Utah. The study assumes a repository capacity of 36,000 metric tons of heavy metal (MTHM) of unreprocessed spent fuel and 36,000 MTHM of commercial high-level reprocessing waste, along with 7,020 canisters of defense high-level reprocessing waste and associated quantities of remote- and contact-handled transuranic waste (TRU). With the exception of TRU, all the waste forms are placed in 300- to 1,000-year-life carbon-steel waste packages in a collected waste handling and packaging facility (WHPF), which is also described. The construction, operation, and decommissioning of the proposed repository is estimated to cost approximately $5.49 billion. Costs include those for the collocated WHPF, engineering, and contingency, but exclude waste form assembly and shipment to the site and waste package fabrication and shipment to the site. These costs reflect the relative average wage rates of the region and the relatively sound nature of the salt at this site. Construction would require an estimated 7.75 years. Engineering factors and costs are not strongly influenced by environmental considerations. 50 refs., 24 figs., 20 tabs

  17. Basic repository environmental assessment design basis, Lavender Canyon site

    International Nuclear Information System (INIS)

    1988-01-01

    This study examines the engineering factors and costs associated with the construction, operation, and decommissioning of a high-level nuclear waste repository in salt in the Paradox Basin in Lavender Canyon, Utah. The study assumes a repository capacity of 36,000 metric tons of heavy metal (MTHM) of unreprocessed spent fuel and 36,000 MTHM of commercial high-level reprocessing waste, along with 7020 canisters of defense high-level reprocessing waste and associated quantities of remote- and contact-handled transuranic waste (TRU). With the exception of TRU, all the waste forms are placed in 300- to 1000-year-life carbon-steel waste packages in a collocated waste handling and packaging facility (WHPF), which is also described. The construction, operation, and decommissioning of the proposed repository is estimated to cost approximately $5.51 billion. Costs include those for the collocated WHPP, engineering, and contingency, but exclude waste form assembly and shipment to the site and waste package fabrication and shipment to the site. These costs reflect the relative average wage rates of the region and the relatively sound nature of the salt at this site. Construction would require an estimated 7.75 years. Engineering factors and costs are not strongly influenced by environmental considerations. 51 refs., 24 figs., 20 tabs

  18. Quality assurance in performance assessments

    International Nuclear Information System (INIS)

    Maul, P.R.; Watkins, B.M.; Salter, P.; Mcleod, R

    1999-01-01

    Following publication of the Site-94 report, SKI wishes to review how Quality Assurance (QA) issues could be treated in future work both in undertaking their own Performance Assessment (PA) calculations and in scrutinising documents supplied by SKB (on planning a repository for spent fuels in Sweden). The aim of this report is to identify the key QA issues and to outline the nature and content of a QA plan which would be suitable for SKI, bearing in mind the requirements and recommendations of relevant standards. Emphasis is on issues which are specific to Performance Assessments for deep repositories for radioactive wastes, but consideration is also given to issues which need to be addressed in all large projects. Given the long time over which the performance of a deep repository system must be evaluated, the demonstration that a repository is likely to perform satisfactorily relies on the use of computer-generated model predictions of system performance. This raises particular QA issues which are generally not encountered in other technical areas (for instance, power station operations). The traceability of the arguments used is a key QA issue, as are conceptual model uncertainty, and code verification and validation; these were all included in the consideration of overall uncertainties in the Site-94 project. Additionally, issues which are particularly relevant to SKI include: How QA in a PA fits in with the general QA procedures of the organisation undertaking the work. The relationship between QA as applied by the regulator and the implementor of a repository development programme. Section 2 introduces the discussion of these issues by reviewing the standards and guidance which are available from national and international organisations. This is followed in Section 3 by a review of specific issues which arise from the Site-94 exercise. An outline procedure for managing QA issues in SKI is put forward as a basis for discussion in Section 4. It is hoped that

  19. Quality assurance in performance assessments

    Energy Technology Data Exchange (ETDEWEB)

    Maul, P.R.; Watkins, B.M.; Salter, P.; Mcleod, R [QuantiSci Ltd, Henley-on-Thames (United Kingdom)

    1999-01-01

    Following publication of the Site-94 report, SKI wishes to review how Quality Assurance (QA) issues could be treated in future work both in undertaking their own Performance Assessment (PA) calculations and in scrutinising documents supplied by SKB (on planning a repository for spent fuels in Sweden). The aim of this report is to identify the key QA issues and to outline the nature and content of a QA plan which would be suitable for SKI, bearing in mind the requirements and recommendations of relevant standards. Emphasis is on issues which are specific to Performance Assessments for deep repositories for radioactive wastes, but consideration is also given to issues which need to be addressed in all large projects. Given the long time over which the performance of a deep repository system must be evaluated, the demonstration that a repository is likely to perform satisfactorily relies on the use of computer-generated model predictions of system performance. This raises particular QA issues which are generally not encountered in other technical areas (for instance, power station operations). The traceability of the arguments used is a key QA issue, as are conceptual model uncertainty, and code verification and validation; these were all included in the consideration of overall uncertainties in the Site-94 project. Additionally, issues which are particularly relevant to SKI include: How QA in a PA fits in with the general QA procedures of the organisation undertaking the work. The relationship between QA as applied by the regulator and the implementor of a repository development programme. Section 2 introduces the discussion of these issues by reviewing the standards and guidance which are available from national and international organisations. This is followed in Section 3 by a review of specific issues which arise from the Site-94 exercise. An outline procedure for managing QA issues in SKI is put forward as a basis for discussion in Section 4. It is hoped that

  20. Radionuclide Transport in Fractured Rock: Numerical Assessment for High Level Waste Repository

    Directory of Open Access Journals (Sweden)

    Claudia Siqueira da Silveira

    2013-01-01

    Full Text Available Deep and stable geological formations with low permeability have been considered for high level waste definitive repository. A common problem is the modeling of radionuclide migration in a fractured medium. Initially, we considered a system consisting of a rock matrix with a single planar fracture in water saturated porous rock. Transport in the fracture is assumed to obey an advection-diffusion equation, while molecular diffusion is considered the dominant mechanism of transport in porous matrix. The partial differential equations describing the movement of radionuclides were discretized by finite difference methods, namely, fully explicit, fully implicit, and Crank-Nicolson schemes. The convective term was discretized by the following numerical schemes: backward differences, centered differences, and forward differences. The model was validated using an analytical solution found in the literature. Finally, we carried out a simulation with relevant spent fuel nuclide data with a system consisting of a horizontal fracture and a vertical fracture for assessing the performance of a hypothetical repository inserted into the host rock. We have analysed the bentonite expanded performance at the beginning of fracture, the quantified radionuclide released from a borehole, and an estimated effective dose to an adult, obtained from ingestion of well water during one year.

  1. 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.

  2. 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)

  3. 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

    important differences between KBS-3H and KBS-3V occur. The description of evolution in this phase addresses in turn the (i), thermal evolution, (ii), groundwater flow and evolution of groundwater composition, (iii), mechanical evolution, (iv), saturation and buffer swelling, (v) evolution of chemical and microbiological conditions and (iv), evolution of the canister surface and interior. A key issue is the local variability of the near-field rock around the KBS-3H repository drifts, which, together with the effects of gas from the corrosion of the supercontainer shells and other steel repository components, results in widely differing saturation times for different drift sections. Nevertheless, even in the tightest drift sections, the buffer is expected to retain its initial water content, and will eventually fully saturate, at which time it is expected to perform its full range of safety functions. In the more distant future, the evolution of conditions at repository depth may be significantly affected by major climate change, and, in particular, by the formation of ice sheets at the ground surface. Key issues, which are common to KBS-3H and KBS-3V, are (i), the possibility that meltwater penetrating to repository depth may lead to some erosion of the buffer, and (ii), the impact of post-glacial earthquakes on canister integrity. The impact of anthropogenic emissions on the magnitude and timing of future major climate changes is a key uncertainty that is also common to the two alternatives. The description of repository evolution provides the basis for the identification of evolution scenarios, an assessment of canister longevity and the analysis of radionuclide release and transport in the event of canister failure. Scenarios involving canister failure and radionuclide release are, however, also identified based on the discussion in this report and in the KBS-3H Process Report. These are initiated, in the first place, by: the presence of an initial, penetrating defect in

  4. 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

    important differences between KBS-3H and KBS-3V occur. The description of evolution in this phase addresses in turn the (i), thermal evolution, (ii), groundwater flow and evolution of groundwater composition, (iii), mechanical evolution, (iv), saturation and buffer swelling, (v) evolution of chemical and microbiological conditions and (vi), evolution of the canister surface and interior. A key issue is the local variability of the near-field rock around the KBS-3H repository drifts, which, together with the effects of gas from the corrosion of the supercontainer shells and other steel repository components, results in widely differing saturation times for different drift sections. Nevertheless, even in the tightest drift sections, the buffer is expected to retain its initial water content, and will eventually fully saturate, at which time it is expected to perform its full range of safety functions. In the more distant future, the evolution of conditions at repository depth may be significantly affected by major climate change, and, in particular, by the formation of ice sheets at the ground surface. Key issues, which are common to KBS-3H and KBS.3V, are (i), the possibility that meltwater penetrating to repository depth may lead to some erosion of the buffer, and (ii), the impact of post-glacial earthquakes on canister integrity. The impact of anthropogenic emissions on the magnitude and timing of future major climate changes is a key uncertainty that is also common to the two alternatives. The description of repository evolution provides the basis for the identification of evolution scenarios, an assessment of canister longevity and the analysis of radionuclide release and transport in the event of canister failure. Scenarios involving canister failure and radionuclide release are, however, also identified based on the discussion in this report and in the KBS-3H Process Report. These are initiated, in the first place, by: . the presence of an initial, penetrating defect

  5. 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

    important differences between KBS-3H and KBS-3V occur. The description of evolution in this phase addresses in turn the (i), thermal evolution, (ii), groundwater flow and evolution of groundwater composition, (iii), mechanical evolution, (iv), saturation and buffer swelling, (v) evolution of chemical and microbiological conditions and (vi), evolution of the canister surface and interior. A key issue is the local variability of the near-field rock around the KBS-3H repository drifts, which, together with the effects of gas from the corrosion of the supercontainer shells and other steel repository components, results in widely differing saturation times for different drift sections. Nevertheless, even in the tightest drift sections, the buffer is expected to retain its initial water content, and will eventually fully saturate, at which time it is expected to perform its full range of safety functions. In the more distant future, the evolution of conditions at repository depth may be significantly affected by major climate change, and, in particular, by the formation of ice sheets at the ground surface. Key issues, which are common to KBS-3H and KBS.3V, are (i), the possibility that meltwater penetrating to repository depth may lead to some erosion of the buffer, and (ii), the impact of post-glacial earthquakes on canister integrity. The impact of anthropogenic emissions on the magnitude and timing of future major climate changes is a key uncertainty that is also common to the two alternatives. The description of repository evolution provides the basis for the identification of evolution scenarios, an assessment of canister longevity and the analysis of radionuclide release and transport in the event of canister failure. Scenarios involving canister failure and radionuclide release are, however, also identified based on the discussion in this report and in the KBS-3H Process Report. These are initiated, in the first place, by: . the presence of an initial, penetrating defect

  6. Assessing the economic and fiscal effects of repository siting

    International Nuclear Information System (INIS)

    Chase, R.A.; Leistritz, F.L.; Halstead, J.M.

    1983-01-01

    The development of high-level nuclear waste repositories can be expected to produce a substantial stimulus to the local economy, but the magnitude and speed of this stimulus may pose planning problems for nearby communities. The chapter is divided into two major sections: 1) a review of the major economic and fiscal effects likely to occur, and 2) a discussion of the methods and techniques which can be used to estimate the magnitude, timing, and distribution of such effects. The six categories of effects described are those on employment patterns and characteristics, income, local trade and service firms, community price structure changes, other basic economic sectors, and public sector costs and revenues

  7. Scientific basis for long-term prediction of waste-form performance under repository conditions

    International Nuclear Information System (INIS)

    Mendel, J.E.

    1982-10-01

    This paper presents an overview of the fundamental principles involved in predicting long-term performance of waste forms by the as-low-as-reasonably-achievable approach. Repository conditions which make up the waste-form environment, the aging of the waste form, the important radionuclides in the waste form, the chemistry of repository fluids, and multicomponent interactions testing were considered in order to describe these principles. The need for confidence limits on the prediction of waste-form performance and ways of achieving a definition of the confidence limits are discussed

  8. Coupled Multi-physical Simulations for the Assessment of Nuclear Waste Repository Concepts: Modeling, Software Development and Simulation

    Science.gov (United States)

    Massmann, J.; Nagel, T.; Bilke, L.; Böttcher, N.; Heusermann, S.; Fischer, T.; Kumar, V.; Schäfers, A.; Shao, H.; Vogel, P.; Wang, W.; Watanabe, N.; Ziefle, G.; Kolditz, O.

    2016-12-01

    As part of the German site selection process for a high-level nuclear waste repository, different repository concepts in the geological candidate formations rock salt, clay stone and crystalline rock are being discussed. An open assessment of these concepts using numerical simulations requires physical models capturing the individual particularities of each rock type and associated geotechnical barrier concept to a comparable level of sophistication. In a joint work group of the Helmholtz Centre for Environmental Research (UFZ) and the German Federal Institute for Geosciences and Natural Resources (BGR), scientists of the UFZ are developing and implementing multiphysical process models while BGR scientists apply them to large scale analyses. The advances in simulation methods for waste repositories are incorporated into the open-source code OpenGeoSys. Here, recent application-driven progress in this context is highlighted. A robust implementation of visco-plasticity with temperature-dependent properties into a framework for the thermo-mechanical analysis of rock salt will be shown. The model enables the simulation of heat transport along with its consequences on the elastic response as well as on primary and secondary creep or the occurrence of dilatancy in the repository near field. Transverse isotropy, non-isothermal hydraulic processes and their coupling to mechanical stresses are taken into account for the analysis of repositories in clay stone. These processes are also considered in the near field analyses of engineered barrier systems, including the swelling/shrinkage of the bentonite material. The temperature-dependent saturation evolution around the heat-emitting waste container is described by different multiphase flow formulations. For all mentioned applications, we illustrate the workflow from model development and implementation, over verification and validation, to repository-scale application simulations using methods of high performance computing.

  9. Melodie: a code for risk assessment of waste repositories in deep geological formations

    International Nuclear Information System (INIS)

    Lewi, J.; Mejon-Goula, M.J.; Cernes, A.

    1988-10-01

    In order to perform the safety evaluation of nuclear waste repositories, a global model, called MELODIE, is currently developed at the CEA/IPSN, in collaboration with order CEA teams and non-CEA like ENSMP (Ecole Nationale Superieure des Mines de Paris). The version now in operation allows to assess the radiological consequences due to a repository located in a granitic formation on a period of several hundred thousands of years. The calculations are based on models which represent the physical and chemical phenomena in connection with: the release of the radionuclides from the waste matrixes and through the engineered barriers; their transfer through the geosphere; their behaviour in the biosphere. Three separate models have been developed for each of these subjects; they are integrated in the code through a modular flexible dataprocessing structure which calls these computational modules with their optimal time step and extracts the data from the data files where they are stored. In addition, a sensitivity and uncertainty analysis algorithm has been implemented into the code. It allows to evaluate the influence of the parameter values on the result and to assess the global uncertainty on it. After a quite general description of MELODIE, the calculations performed with it in the PAGIS (CCE) exercise: global dose calculations and ranking of the most important parameters through the sensitivity analysis, are presented. The studies performed only with the geosphere module of MELODIE (METIS), especially the participation to the HYDROCOIN (OECD/NEA) exercise, are also noticed. In addition, the main future development axes of MELODIE are outlined

  10. 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.

  11. 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

  12. 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)

  13. The application of geological computer modelling systems to the characterisation and assessment of radioactive waste repositories

    International Nuclear Information System (INIS)

    White, M.J.; Del Olmo, C.

    1996-01-01

    The deep disposal of radioactive waste requires the collection and analysis of large amounts of geological data. These data give information on the geological and hydrogeological setting of repositories and research sites, including the geological structure and the nature of the groundwater. The collection of these data is required in order to develop an understanding of the geology and the geological evolution of sites and to provide quantitative information for performance assessments. An integrated approach to the interpretation and provision of these data is proposed in this paper, via the use of computer systems, here termed geological modelling systems. Geological modelling systems are families of software programmes which allow the incorporation of site investigation data into integrated 3D models of sub-surface geology

  14. The link between performance assessment and quality of data

    International Nuclear Information System (INIS)

    Csullog, G.W.

    1990-11-01

    Performance assessments, using both individual and cumulative inventories, produce estimates of the risk or dose consequences for repository concepts. Unacceptable risk or dose for a given contaminant identify it as a problem contaminant. While risk or dose consequences are common outputs of safety assessments, results can also be reported as estimates of the safe inventory limit for each contaminant in a repository scenario. Once problem contaminants are identified, it may be useful to: validate the data used in the performance assessment; restrict contaminant inventories in the repository; or, improve the containment capability of the repository (including improved packaging). A key point is that performance assessments define the quality of data to be collected. If a contaminant's inventory will be far below its estimated safe inventory limit for a repository, then it is reasonable to accept inventory estimates that have high variances and low confidence. For contaminants that have a significant impact upon performance assessments, efforts must be made to provide the best inventory (source term) estimates practicable. One important goal of characterization programs should be to establish the adequacy of data used in performance assessments. There is little value in striving to improve confidence in data for some waste streams if they are assessed to have a minor impact upon repository performance. Streams that have a major impact upon performance deserve the time, attention and money to be characterized adequately. Waste acceptance must be linked to the quality of characterization data provided by generators. This link sets compliance monitoring requirements

  15. Performance of high level waste forms and engineered barriers under repository conditions

    International Nuclear Information System (INIS)

    1991-02-01

    The IAEA initiated in 1977 a co-ordinated research programme on the ''Evaluation of Solidified High-Level Waste Forms'' which was terminated in 1983. As there was a continuing need for international collaboration in research on solidified high-level waste form and spent fuel, the IAEA initiated a new programme in 1984. The new programme, besides including spent fuel and SYNROC, also placed greater emphasis on the effect of the engineered barriers of future repositories on the properties of the waste form. These engineered barriers included containers, overpacks, buffer and backfill materials etc. as components of the ''near-field'' of the repository. The Co-ordinated Research Programme on the Performance of High-Level Waste Forms and Engineered Barriers Under Repository Conditions had the objectives of promoting the exchange of information on the experience gained by different Member States in experimental performance data and technical model evaluation of solidified high level waste forms, components of the waste package and the complete waste management system under conditions relevant to final repository disposal. The programme includes studies on both irradiated spent fuel and glass and ceramic forms as the final solidified waste forms. The following topics were discussed: Leaching of vitrified high-level wastes, modelling of glass behaviour in clay, salt and granite repositories, environmental impacts of radionuclide release, synroc use for high--level waste solidification, leachate-rock interactions, spent fuel disposal in deep geologic repositories and radionuclide release mechanisms from various fuel types, radiolysis and selective leaching correlated with matrix alteration. Refs, figs and tabs

  16. 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.

  17. 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

  18. 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

  19. 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

  20. 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

  1. 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

  2. Fuzzy distributions in probabilistic environmental impact assessment: application to a high-level waste repository

    International Nuclear Information System (INIS)

    Datta, D.; Joshi, M.L.

    2006-01-01

    Environmental modeling with a satisfaction levels of the end user in relation to a defined parameter coupled with imprecision that stems from the field data is a key issue. In the context of this issue success of possibility theory based on fuzzy sets has high visibility in comparison with conventional probability theory. Environmental impact assessments of a high level waste repository is focused using the new approach because the problems under consideration includes a number of qualitative uncertainties at different levels, apart from being quite complex; decision-maker's need to have a transparent assessment result that will enable him to understand underlying assumptions and to judge resulting doses. Fuzzy distributions have been tried to resolve the issues related to the safety of environment from the waste repository. Paper describes the details of fuzzy distribution, fuzzy logic and its possible application to deal the qualitative and quantitative uncertainty in connection with waste repository. (author)

  3. Performance assessment modeling of pyrometallurgical process wasteforms

    International Nuclear Information System (INIS)

    Nutt, W.M.; Hill, R.N.; Bullen, D.B.

    1995-01-01

    Performance assessment analyses have been completed to estimate the behavior of high-level nuclear wasteforms generated from the pyrometallurgical processing of liquid metal reactor (LMR) and light water reactor (LWR) spent nuclear fuel. Waste emplaced in the proposed repository at Yucca Mountain is investigated as the basis for the study. The resulting cumulative actinide and fission product releases to the accessible environment within a 100,000 year period from the various pyrometallurgical process wasteforms are compared to those of directly disposed LWR spent fuel using the same total repository system model. The impact of differing radionuclide transport models on the overall release characteristics is investigated

  4. Preliminary radiological assessments of near-surface low-level radioactive waste repositories

    International Nuclear Information System (INIS)

    Sumerling, T.J.; Nancarrow, D.J.

    1988-08-01

    This report summarises preliminary assessments of post-closure radiological impact of LLW repositories at four sites previously under investigation by UK Nirex Ltd. The objectives of the assessments were: to demonstrate a methodology for site specific assessments, to identify important information requirements for detailed assessments; to identify methodological and research requirements. Doses and risks due to groundwater pathways, human intrusion, gaseous release and natural environmental change are estimated. (author)

  5. 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

  6. 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

  7. 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

  8. Repository simulation tests

    International Nuclear Information System (INIS)

    Wicks, G.G.; Bibler, N.E.; Jantzen, C.M.; Plodinec, M.J.

    1984-01-01

    The repository simulation experiments described in this paper are designed to assess the performance of SRP waste glass under the most realistic repository conditions that can be obtained in the laboratory. These tests simulate the repository environment as closely as possible and introduce systematically the variability of the geology, groundwater chemistry, and waste package components during the leaching of the waste glass. The tests evaluate waste form performance under site-specific conditions, which differ for each of the geologic repositories under consideration. Data from these experiments will aid in the development of a realistic source term that can describe the release of radionuclides from SRP waste glass as a component of proposed waste packages. Hence, this information can be useful to optimize waste package design for SRP waste glass and to provide data for predicting long-term performance and subsequent conformance to regulations. The repository simulation tests also help to bridge the gap in interpreting results derived from tests performed under the control of the laboratory to the uncertainity and variability of field tests. In these experiments, site-specific repository components and conditions are emphasized and only the site specific materials contact the waste forms. An important feature of these tests is that both actual and simulated waste glasses are tested identically. 7 figures, 2 tables

  9. Viability Assessment of a Repository at Yucca Mountain. Volume 4: License Application Plan and Costs

    Energy Technology Data Exchange (ETDEWEB)

    None

    1998-12-01

    Volume 4 provides the DOE plan and cost estimate for the remaining work necessary to proceed from completing this VA to submitting an LA to NRC. This work includes preparing an EIS and evaluating the suitability of the site. Both items are necessary components of the documentation required to support a decision in 2001 by the Secretary of Energy on whether or not to recommend that the President approve the site for development as a repository. If the President recommends the site to Congress and the site designation becomes effective, then DOE will submit the LA to NRC in 2002 for authorization to construct the repository. The work described in Volume 4 constitutes the last step in the characterization of the Yucca Mountain site and the design and evaluation of the performance of a repository system in the geologic setting of this site. The plans in this volume for the next 4 years' work are based on the results of the previous 15 years' work, as reported in Volumes 1, 2, and 3 of this VA. Volume 1 summarizes what DOE has learned to date about the Yucca Mountain site. Volume 2 describes the current, reference repository design, several design options that might enhance the performance of the reference design, and several alternative designs that represent substantial departures from the reference design. Volume 2 also summarizes the results of tests of candidate materials for waste packages and for support of the tunnels into which waste would be emplaced. Volume 3 provides the results of the latest performance assessments undertaken to evaluate the performance of the design in the geologic setting of Yucca Mountain. The results described in Volumes 1, 2, and 3 provide the basis for identifying and prioritizing the work described in this volume. DOE believes that the planned work, together with the results of previous work, will be sufficient to support a site suitability evaluation for site recommendation and, if the site is recommended and designated, a

  10. 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.)

  11. 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)

  12. 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)

  13. Hydrologic Monitoring in the Deep Subsurface to Support Repository Performance

    Science.gov (United States)

    Hubbell, J. M.; Heath, G. L.; Scott, C. L.

    2007-12-01

    The INL has installed and operated several vadose and ground water monitoring systems in arid and humid sites to depths of about 200m. Some of these systems have been in continuous operation for over 12 years. It is important that the systems be physically robust, simple, yet versatile enough that it can operate for extended time periods with little or no maintenance. Monitoring instruments are frequently installed and run to characterize the site, collect data during site operation, and continue to run for long-term stewardship, necessitating sensors that can be maintained or serviced. Sensors are carefully chosen based on the perceived data requirements over the life of the site. An emphasis is given on direct measurements such as tensiometers (portable and advanced), neutron probe, drain gauge, temperature, wells or sampling for fluids and gases. Other complementary data can include using TDR/capacitance, radiation detectors, and larger scale geophysical techniques (3-d resistivity and EM) for volumetric measurements. Commercially available instruments may have to be modified for their use at greater depths, to allow multiple instruments in a single borehole or to perform the intended monitoring function. Access tubes (some open at the bottom) can be placed to allow insertion of multiple sensors (radiation, neutron and portable sensors/samplers), future drilling/sampling and to install new instruments at a later time. The installation techniques and backfill materials must be chosen and the measurement technique tested to ensure representative data collection for the parameters of interest. The data collection system can be linked to climatic data (precipitation, barometric pressure, snow depth, runoff, surface water sources) that may influence the site's subsurface hydrology. The instruments are then connected to a real-time automated data collection system that collect, stores, and provides access to the data. These systems have been developed that allow easy

  14. Thermal Performance Analysis of a Geologic Borehole Repository

    Energy Technology Data Exchange (ETDEWEB)

    Reagin, Lauren [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-08-16

    The Brazilian Nuclear Research Institute (IPEN) proposed a design for the disposal of Disused Sealed Radioactive Sources (DSRS) based on the IAEA Borehole Disposal of Sealed Radioactive Sources (BOSS) design that would allow the entirety of Brazil’s inventory of DSRS to be disposed in a single borehole. The proposed IPEN design allows for 170 waste packages (WPs) containing DSRS (such as Co-60 and Cs-137) to be stacked on top of each other inside the borehole. The primary objective of this work was to evaluate the thermal performance of a conservative approach to the IPEN proposal with the equivalent of two WPs and two different inside configurations using Co-60 as the radioactive heat source. The current WP configuration (heterogeneous) for the IPEN proposal has 60% of the WP volume being occupied by a nuclear radioactive heat source and the remaining 40% as vacant space. The second configuration (homogeneous) considered for this project was a homogeneous case where 100% of the WP volume was occupied by a nuclear radioactive heat source. The computational models for the thermal analyses of the WP configurations with the Co-60 heat source considered three different cooling mechanisms (conduction, radiation, and convection) and the effect of mesh size on the results from the thermal analysis. The results of the analyses yielded maximum temperatures inside the WPs for both of the WP configurations and various mesh sizes. The heterogeneous WP considered the cooling mechanisms of conduction, convection, and radiation. The temperature results from the heterogeneous WP analysis suggest that the model is cooled predominantly by conduction with effect of radiation and natural convection on cooling being negligible. From the thermal analysis comparing the two WP configurations, the results suggest that either WP configuration could be used for the design. The mesh sensitivity results verify the meshes used, and results obtained from the thermal analyses were close to

  15. 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)

  16. Third report, development of site suitability and design performance data base for a high level nuclear waste repository

    International Nuclear Information System (INIS)

    1978-05-01

    This report presents the results of mining and geotechnical studies provided as support for the development of a technical data base suitable for the generation of a regulatory framework governing high-level nuclear waste disposal: flow path models for basalts, salt domes, and crystalline rocks; simplistic model evaluations; loss of administrative control; repository conceptual designs; repository design considerations; and design performance criteria

  17. In situ experiments on the performance of near-field for nuclear waste repository at KURT

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Won-Jin, E-mail: wjcho@kaeri.re.kr [Korea Atomic Energy Research Institute, P.O. Box 105, Yuseong, Daejeon 305-600 (Korea, Republic of); Kim, Jin-Sub; Lee, Changsoo [Korea Atomic Energy Research Institute, P.O. Box 105, Yuseong, Daejeon 305-600 (Korea, Republic of); Kwon, Sangki [Inha University, Department of Energy Resources Engineering, 253 Yonghyun-dong, Nam-gu, Incheon 402-751 (Korea, Republic of); Choi, Jong-Won [Korea Atomic Energy Research Institute, P.O. Box 105, Yuseong, Daejeon 305-600 (Korea, Republic of)

    2012-11-15

    Highlights: Black-Right-Pointing-Pointer Results of in situ experiments on the near-field of a repository are summarized. Black-Right-Pointing-Pointer In BHT, the rock temperatures at 0.3 m from heater showed 40-50 Degree-Sign C increase. Black-Right-Pointing-Pointer EDZ size measured from the in situ test was in the range of 0.6-1.8 m. Black-Right-Pointing-Pointer Maximum errors in estimating the location of AE source were 0.6-1.0 m. Black-Right-Pointing-Pointer Permeability in the EDZ increased up to 2 orders compared with the intact rock. - Abstract: To obtain the information on the design and performance assessment of a geological repository for nuclear waste, several in situ experiments on the performance of the near-field have been conducted for 5 years in the small-scale underground research laboratory, KURT. This paper summarizes the results from the in situ experiments. In the borehole heater test, the rock temperature at 0.3 m distance from the heater hole with 90 Degree-Sign C showed a 40 Degree-Sign C increase over initial temperature. After heating with 120 Degree-Sign C, the rock temperature at 0.3 m distance from the heater hole increased up to 50 Degree-Sign C over initial temperature. The EDZ size measured from the in situ test was in the range of 0.6-1.8 m, and was higher than that from the laboratory tests, which was estimated to be around 1.1-1.5 m. The maximum errors in estimating the location of acoustic emission source were 1.0 m in EDZ and 0.6 m in an intact rock zone, respectively. The damping ratios of the EDZ and intact rock were 0.091 and 0.005, respectively. The permeability of rock increased with decreasing distance from the tunnel wall because of the EDZ. The permeability in the EDZ seems to be increased up to 2 orders compared with that in the intact rock.

  18. Applying insights from repository safety assessments to evaluating impacts of partitioning and transmutation

    International Nuclear Information System (INIS)

    Nutt, W. Mark; Swift, Peter N.

    2010-01-01

    Published analyses of geologic repositories indicate potential for excellent long-term performance for a range of disposal concepts. Estimates of peak dose may be dominated by different radionuclides in different disposal concepts. Thermal loading issues can be addressed by design and operational choices. Impact of waste form lifetime on estimates of peak dose varies for different disposal concepts.

  19. The role of performance assessment in radioactive waste disposal

    International Nuclear Information System (INIS)

    Stenhouse, M.J.

    1998-01-01

    Performance assessment has many applications in the field of radioactive waste management, none more important than demonstrating the suitability of a particular repository system for waste disposal. The role of performance assessment in radioactive waste disposal is discussed with reference to assessments performed in civilian waste management programmes. The process is, however, relevant, and may be applied directly to the disposal of defence-related wastes. When used in an open and transparent manner, performance assessment is a powerful methodology not only for convincing the authorities of the safety of a disposal concept, but also for gaining the wider acceptance of the general public for repository siting. 26 refs

  20. The feasibility of Backfilling a Repository of Spent Fuel: An Assessment of Recent Developments by SKB

    Energy Technology Data Exchange (ETDEWEB)

    Bennett, David (TerraSalus Limited, Rutland (United Kingdom))

    2010-12-15

    In the Review Statement and Evaluation of SKB's RDandD programme 2007 (SKI Report 2008:48E), the former Swedish Nuclear Power Inspectorate (SKI) commented that considerable work remained to be done for knowledge of both practical management issues on backfilling and analysis of long-term backfill evolution to reach the same level as for the canister and the buffer. SKI considered that the backfill material had not been thoroughly reported in the RDandD programme. More concrete plans were also needed relating to large-scale demonstration experiments to investigate the performance of the backfill in as realistic conditions as possible. In the spring of 2009, noting that SKB had changed its concept for backfilling several times over the last few years, and after having visited SKB's most recent backfilling trials at Aespoe, both SSM and SSM's expert group BRITE had strong concerns regarding SKB's programme for backfilling the repository tunnels. Although the BRITE expert group has been keeping a watching brief over SKB's development work on backfilling, SSM has not undertaken a systematic assessment of SKB's work in this area since the SR-Can Safety Report was reviewed in 2006. Dr David Bennett, a member and secretary of the BRITE expert group, was asked to do such an assessment. This report describes the assessment results

  1. The feasibility of Backfilling a Repository of Spent Fuel: An Assessment of Recent Developments by SKB

    International Nuclear Information System (INIS)

    Bennett, David

    2010-12-01

    In the Review Statement and Evaluation of SKB's RDandD programme 2007 (SKI Report 2008:48E), the former Swedish Nuclear Power Inspectorate (SKI) commented that considerable work remained to be done for knowledge of both practical management issues on backfilling and analysis of long-term backfill evolution to reach the same level as for the canister and the buffer. SKI considered that the backfill material had not been thoroughly reported in the RDandD programme. More concrete plans were also needed relating to large-scale demonstration experiments to investigate the performance of the backfill in as realistic conditions as possible. In the spring of 2009, noting that SKB had changed its concept for backfilling several times over the last few years, and after having visited SKB's most recent backfilling trials at Aespoe, both SSM and SSM's expert group BRITE had strong concerns regarding SKB's programme for backfilling the repository tunnels. Although the BRITE expert group has been keeping a watching brief over SKB's development work on backfilling, SSM has not undertaken a systematic assessment of SKB's work in this area since the SR-Can Safety Report was reviewed in 2006. Dr David Bennett, a member and secretary of the BRITE expert group, was asked to do such an assessment. This report describes the assessment results

  2. Salt site performance assessment activities

    International Nuclear Information System (INIS)

    Kircher, J.F.; Gupta, S.K.

    1983-01-01

    During this year the first selection of the tools (codes) for performance assessments of potential salt sites have been tentatively selected and documented; the emphasis has shifted from code development to applications. During this period prior to detailed characterization of a salt site, the focus is on bounding calculations, sensitivity and with the data available. The development and application of improved methods for sensitivity and uncertainty analysis is a focus for the coming years activities and the subject of a following paper in these proceedings. Although the assessments to date are preliminary and based on admittedly scant data, the results indicate that suitable salt sites can be identified and repository subsystems designed which will meet the established criteria for protecting the health and safety of the public. 36 references, 5 figures, 2 tables

  3. Preliminary radiological assessments of low-level waste repositories

    International Nuclear Information System (INIS)

    Nancarrow, D.J.; Sumerling, T.J.; Ashton, J.

    1988-06-01

    Preliminary assessments of the post-closure radiological impact from the disposal of low-level radioactive wastes in shallow engineered facilities at four sites are presented. This provides a framework to practice and refine a methodology that could be used, on behalf of the Department, for independent assessment of any similar proposal from Nirex. Information and methodological improvements that would be required are identified. (author)

  4. Identifying significant uncertainties in thermally dependent processes for repository performance analysis

    International Nuclear Information System (INIS)

    Gansemer, J.D.; Lamont, A.

    1994-01-01

    In order to study the performance of the potential Yucca Mountain Nuclear Waste Repository, scientific investigations are being conducted to reduce the uncertainty about process models and system parameters. This paper is intended to demonstrate a method for determining a strategy for the cost effective management of these investigations. It is not meant to be a complete study of all processes and interactions, but does outline a method which can be applied to more in-depth investigations

  5. Parametric study of geohydrologic performance characteristics for nuclear-waste repositories

    International Nuclear Information System (INIS)

    Bailey, C.E.; Marine, I.W.

    1981-01-01

    Purpose of this study was to present geohydrologic information in graphical form covering a wide range of parameters to aid in determining site specifications based on functional criteria. Graphs of the major performance characteristics that influence the transport of radionuclides from a repository to the biosphere by groundwater were developed. The major characteristics addressed are radioactive decay, leach rate, hydraulic conductivity, porosity, groundwater gradient, hydrodynamic dispersion, ion exchange, and distance to the biosphere

  6. An overview of performance assessment methodology

    International Nuclear Information System (INIS)

    Hongnian Jow

    2010-01-01

    The definition of performance assessment (PA) within the context of a geologic repository program is a post-closure safety assessment; a system analysis of hazards associated with the facility and the ability of the site and the design of the facility to provide for the safety functions. For the last few decades, PA methodology bas been developed and applied to different waste disposal programs around the world. PA has been used in the safety analyses for waste disposal repositories for low-level waste, intermediate level waste, and high-level waste including spent nuclear fuels. This paper provides an overview of the performance assessment methodology and gives examples of its applications for the Yucca Mountain Project. (authors)

  7. 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

  8. 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

  9. Interim performance specifications for conceptual waste-package designs for geologic isolation in salt repositories

    International Nuclear Information System (INIS)

    1983-06-01

    The interim performance specifications and data requirements presented apply to conceptual waste package designs for all waste forms which will be isolated in salt geologic repositories. The waste package performance specifications and data requirements respond to the waste package performance criteria. Subject areas treated include: containment and controlled release, operational period safety, criticality control, identification, and waste package performance testing requirements. This document was generated for use in the development of conceptual waste package designs in salt. It will be revised as additional data, analyses, and regulatory requirements become available

  10. Key radionuclides and parameters that determine performance of geologic repositories for high-level radioactive wastes

    International Nuclear Information System (INIS)

    Joonhong Ahn; Atsuyuki Suzuki

    1993-01-01

    This paper presents results of a mathematical analysis for performance of the engineered barriers of high-level radioactive waste repositories. The main body of the mathematical model developed in this study is mass transport of actinides in a bentonite region. In an analysis of actinide transport, radioactive decay chain and effects of low solubilities must be taken into account. In many previous models for mass transport in engineered barriers including radioactive decay chain, however, boundary conditions at the interface between the waste form and the bentonite region cannot be determined flexibly. In some models, solubility-limited boundary condition is assumed for all the members in a chain. In order to investigate what are key radionuclides and parameters that control performance of engineered barriers of a geologic repository, we must evaluate mass transport with the source boundary condition determined by a detailed analysis on mass transfer at the boundary. In this study, we developed a mathematical model, which can determine whether the inner boundary condition is solubility-limited or congruent release, based on a mathematical analysis for mass transfer at the glass dissolution location, and how long the solubility-limited boundary condition applies. Based on the mathematical model, we point out radionuclides and parameters that have primary influences on the performance of a repository, and investigate a reasonable strategy for coupling geologic disposal and partitioning of those key radionuclides from the standpoint of reducing hazard of geologic disposal. (authors). 4 tabs., 2 figs., 8 refs

  11. Microbial processes relevant for the long-term performance of radioactive waste repositories in clays

    International Nuclear Information System (INIS)

    Meleshyn, Artur

    2012-01-01

    Document available in extended abstract form only. A number of investigations on occurrence and viability of microbes in compacted clays have been aimed at studying possible microbial effects on long-term performance of a deep geological repository (DGR) for high-level radioactive waste (HLW) and spent nuclear fuel (SF). Compacted clays are considered in current DGR designs either as a buffer material or as a host rock. The primary purpose of the present work was to qualitatively evaluate the relevance of microbial activity for the long-term performance of a DGR and to identify which safety-relevant processes and properties can be potentially influenced by this activity. The present analysis identified eight clay properties essential for maintaining safety functions of containment and retardation of the disposal system - swelling pressure, specific surface area, cation exchange capacity, anion sorption capacity, porosity, permeability, fluid pressure, plasticity - which can potentially be influenced by microbial processes in clay buffer and Clay-stone within a DGR for HLW/SF. Iron(III)- and sulphate-reducing, fermentative, methane-producing and oxidizing microbes can be considered to be present in any clay formation. Each habitat includes a massive number of microbial niches with perhaps only a small proportion of the species being metabolically active at the habitat's conditions, the remainder becoming not extinct. Moreover, clays contain electron donors and electron acceptors in amounts sufficient for these microbes to remain active during very long periods of time. Additional sources of electron donors or electron acceptors will inevitably be added to the repository system as a result of DGR excavation, placement of radioactive waste as well as backfilling and sealing of the DGR. In no case should the potential impact of microbes be underestimated based on a possible argument of comparably low biomass of the microbes in contact with metal surfaces or dissolved

  12. Assessment of human intrusion and future human actions – Example from the Swedish LILW repository

    International Nuclear Information System (INIS)

    Andersson, Eva; Hjerpe, T.; Smith, G.

    2016-01-01

    Recommendations for FHA assessments: • The assumption of only considering inadvertent intrusion and using todays practice and technology is a key tool to make the assessment understandable. • Even after siting has been done, the placement of the repository, should be discussed in FHA scenarios, e.g. placement away from resources, discussion of erosion, and land uplift. • FEPs a good tool, good for constructing relevant scenarios and also for communication of the FHA assessment. • A few stylized scenarios are enough, good to include a drilling scenario as it is often used by different organizations. • Precise probabilities are hard to defend, avoid if possible and instead use the low likelihood in communication of the FHA results. • Communication of FHA scenarios is a tool to build confidence in the robustness of repository systems

  13. 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

  14. 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

  15. 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

  16. Environmental assessment: Reference repository location, Hanford Site, Washington

    International Nuclear Information System (INIS)

    1986-05-01

    This appendix responds to the issues raised by Federal, State, and local governments, affected Indian Tribes, private citizens, and other organizations on the draft environmental assessment (EA) that was prepared pursuant to Section 112 of the Nuclear Waste Policy Act of 1982. In addition to presenting the issues raised in the comments and the responses, it describes where changes were made in the final EA. 535 refs., 13 figs., 16 tabs

  17. Collembase: a repository for springtail genomics and soil quality assessment

    Directory of Open Access Journals (Sweden)

    Klein-Lankhorst Rene M

    2007-09-01

    Full Text Available Abstract Background Environmental quality assessment is traditionally based on responses of reproduction and survival of indicator organisms. For soil assessment the springtail Folsomia candida (Collembola is an accepted standard test organism. We argue that environmental quality assessment using gene expression profiles of indicator organisms exposed to test substrates is more sensitive, more toxicant specific and significantly faster than current risk assessment methods. To apply this species as a genomic model for soil quality testing we conducted an EST sequencing project and developed an online database. Description Collembase is a web-accessible database comprising springtail (F. candida genomic data. Presently, the database contains information on 8686 ESTs that are assembled into 5952 unique gene objects. Of those gene objects ~40% showed homology to other protein sequences available in GenBank (blastx analysis; non-redundant (nr database; expect-value -5. Software was applied to infer protein sequences. The putative peptides, which had an average length of 115 amino-acids (ranging between 23 and 440 were annotated with Gene Ontology (GO terms. In total 1025 peptides (~17% of the gene objects were assigned at least one GO term (expect-value -25. Within Collembase searches can be conducted based on BLAST and GO annotation, cluster name or using a BLAST server. The system furthermore enables easy sequence retrieval for functional genomic and Quantitative-PCR experiments. Sequences are submitted to GenBank (Accession numbers: EV473060 – EV481745. Conclusion Collembase http://www.collembase.org is a resource of sequence data on the springtail F. candida. The information within the database will be linked to a custom made microarray, based on the Agilent platform, which can be applied for soil quality testing. In addition, Collembase supplies information that is valuable for related scientific disciplines such as molecular ecology

  18. Environmental assessment: Reference repository location, Hanford Site, Washington (US)

    Energy Technology Data Exchange (ETDEWEB)

    none,

    1986-05-01

    This appendix responds to the issues raised by Federal, State, and local governments, affected Indian Tribes, private citizens, and other organizations on the draft environmental assessment (EA) that was prepared pursuant to Section 112 of the Nuclear Waste Policy Act of 1982. In addition to presenting the issues raised in the comments and the responses, it describes where changes were made in the final EA. 535 refs., 13 figs., 16 tabs.

  19. Safety performance of a near surface repository subject to a fuel burning

    International Nuclear Information System (INIS)

    Stefanini, Lorenzo; Frano, Rosa Lo; Forasassi, Giuseppe

    2015-01-01

    This study aims to investigate the performances of a near surface repository subject to fuel burning occurring simultaneously or subsequently to a large commercial aircraft impact. Specifically the thermal effects caused by a Boeing-747 crushing (considered like “beyond design basis accident”) are studied. An important part of this study is the analysis of the possible (thermo-mechanical) degradation effects, as dehydration, degasification, pressurization, etc. that the concrete may undergo, particularly in the case of prolonged fire, and of the resistance of structure itself in this condition. Conservative assumptions and restrictions have been made with regard to the fire scenario, the maximum temperature of which is calculated on the basis of the fuel airplane amount, the normal impact, the variation of the material properties along with the temperature as well the damaging phenomena of concrete. The airplane impact load, calculated with the Riera approach, and the maximum temperature, reached during the fuel combustion, are used as input (boundary condition) in the numerical simulations performed by MARC© code. The obtained results showed that a repository wall thickness, ranging from 0.6 to 0.9 m, is not sufficient to prevent the local penetration of wall. To reduce the computational cost, the analyses have been made only on a half part of the structure, highlighting the dominance of thermal effects. Despite the ongoing concrete degradation phenomena, the overall integrity of the repository seemed to be guaranteed as well as the containment and the confinement of radioactive waste. (author)

  20. Assessment of the long-term safety of repositories. Scientific basis

    International Nuclear Information System (INIS)

    Noseck, Ulrich; Becker, Dirk; Fahrenholz, Christine

    2008-12-01

    The project contributed to increase the scientific knowledge on the long-term safety assessment and the safety cases of a radioactive waste repository. International guidelines and more recent safety cases from other countries were evaluated. The feasibility study of the three safety indicators ''individual dose rate'', ''radiotoxicity concentration in the biosphere water'' and ''radiotoxicity flux from the geosphere'' showed that due to the independently derived corresponding reference values these indicators describe three different safety statements. The combination of the three values can give a stronger argument for the safety of the repository system. Another important methodological aspect of the safety cases is the definition and selection of scenarios, one of these the human intrusion scenario. Various human intrusion scenarios are considered in the different nations, which differ significantly with respect to type and time scale, the exposition type and exposition pathway. Further progress has been achieved in how to treat human intrusion scenarios in a German post-closure safety case. Another port of the project dealt with the impact of specific geochemical processes on the long-term safety of the repository. The impact of climate changes on the long-term safety of a radioactive waste repository in rock salt was investigated with respect to processes in the overburden and the biosphere where highest impact is expected. Sofa simplified models and only discrete climate estates have been considered

  1. Properties of altered soils by alkaline solution: contribution in the performance evaluation of repositories

    International Nuclear Information System (INIS)

    Calabria, Jaqueline Alves de Almeida

    2015-01-01

    The radioactive wastes from nuclear technology applications must be properly disposed in a repository, during the necessary time to ensure the human and the environment protection. The surface systems are largely considered for disposal of low and intermediate level radioactive waste, but generally require the use of engineering barriers to control the radionuclides release. An important engineering barrier is the backfill which is situated between the package and structural material, and has the functions of reducing the water infiltration and to promote the retardation of the radionuclide migration, among others. Therefore, the material to be used as backfill such as clays, cement, soils, rocks, must has good sorption, permeability and mechanical properties. The selection of the material will depend also largely on the material availability and installation design. The concrete is also used in the construction of repository, and its interaction with water induces its degradation, resulting in a high pH solution. This solution interacts with the backfilling materials promoting mineralogical alterations that results in significant changes in their key properties and performance as safety component of the repository. In this work, five Brazilian soils of Minas Gerais state, selected according to their generic characteristics along with information from Sistema Brasileiro de Classificacao de Solos (SiBCS) were investigated concerning their potential use as backfilling material in a superficial repository by the determination of retention capacity for cesium and iodine. Sorption-related parameters, used in the performance assessment of the soils, were obtained from experimental data fitting to different sorption isotherms models. The soil that showed the best sorption of Cs, was a clay soil that presented distribution coefficient (K_d) of 90.5 mL.g"-"1 and maximum sorption capacity (Q_m_a_x) of 18.372 mg.g"-"1. Regarding the iodine, the sorption was very low for all

  2. Backfilling and closure of the deep repository. Assessment of backfill concepts

    International Nuclear Information System (INIS)

    Gunnarsson, David; Boergesson, Lennart; Keto, Paula; Tolppanen, Pasi; Hansen, Johanna

    2004-06-01

    This report presents the results from work made in Phase 1 of the joint SKB-Posiva project 'Backfilling and Closure of the Deep Repository' aiming at selecting and developing materials and techniques for backfilling and closure of a KBS-3 type repository for spent nuclear fuel. The aim of phase 1, performed as a desk study, was to describe the potential of the suggested backfill concepts in terms of meeting SKB and Posiva requirements, select the most promising ones for further investigation, and to describe methods that can be used for determining the performance of the concepts. The backfilling concepts described in this report differ from each other with respect to backfill materials and installation techniques. The concepts studied are the following: Concept A: Compaction of a mixture of bentonite and crushed rock in the tunnel. Concept B: Compaction of natural clay with swelling ability in the tunnel. Concept C: Compaction of non-swelling soil type in the tunnel combined with application of pre-compacted bentonite blocks at the roof. Concept D: Placement of pre-compacted blocks; a number of materials are considered. Concept E: Combination of sections consisting of a) crushed rock compacted in the tunnel and b) pre-compacted bentonite blocks. The bentonite sections are installed regularly above every disposal hole. Concept F: Combination of sections consisting of a) crushed rock compacted in the tunnel and b) pre-compacted bentonite blocks. The distance between the bentonite sections is adapted to the local geology and hydrology.The assessment of the concepts is based on performance requirements set for the backfill in the deposition tunnels for providing a stable and safe environment for the bentonite buffer and canister for the repository service time. In order to do this, the backfill should follow certain guidelines, 'design criteria' concerning compressibility, hydraulic conductivity, swelling ability, long-term stability, effects on the barriers and

  3. Backfilling and closure of the deep repository. Assessment of backfill concepts

    Energy Technology Data Exchange (ETDEWEB)

    Gunnarsson, David; Boergesson, Lennart [Clay Technology AB, Lund (Sweden); Keto, Paula [Saanio Riekkola Oy (Finland); Tolppanen, Pasi [Jaakko Poeyry Infra (Finland); Hansen, Johanna [Posiva Oy, Helsinki (Finland)

    2004-06-01

    This report presents the results from work made in Phase 1 of the joint SKB-Posiva project 'Backfilling and Closure of the Deep Repository' aiming at selecting and developing materials and techniques for backfilling and closure of a KBS-3 type repository for spent nuclear fuel. The aim of phase 1, performed as a desk study, was to describe the potential of the suggested backfill concepts in terms of meeting SKB and Posiva requirements, select the most promising ones for further investigation, and to describe methods that can be used for determining the performance of the concepts. The backfilling concepts described in this report differ from each other with respect to backfill materials and installation techniques. The concepts studied are the following: Concept A: Compaction of a mixture of bentonite and crushed rock in the tunnel. Concept B: Compaction of natural clay with swelling ability in the tunnel. Concept C: Compaction of non-swelling soil type in the tunnel combined with application of pre-compacted bentonite blocks at the roof. Concept D: Placement of pre-compacted blocks; a number of materials are considered. Concept E: Combination of sections consisting of a) crushed rock compacted in the tunnel and b) pre-compacted bentonite blocks. The bentonite sections are installed regularly above every disposal hole. Concept F: Combination of sections consisting of a) crushed rock compacted in the tunnel and b) pre-compacted bentonite blocks. The distance between the bentonite sections is adapted to the local geology and hydrology.The assessment of the concepts is based on performance requirements set for the backfill in the deposition tunnels for providing a stable and safe environment for the bentonite buffer and canister for the repository service time. In order to do this, the backfill should follow certain guidelines, 'design criteria' concerning compressibility, hydraulic conductivity, swelling ability, long-term stability, effects on

  4. Technology overview of mined repositories

    International Nuclear Information System (INIS)

    Gimera, R.; Thirumalai, K.

    1982-01-01

    Mined repositories present an environmentally viable option for permanent disposal of nuclear waste. This paper reviews the state-of-the-art mining technologies and identifies technological issues and developments necessary to mine a repository in basalt. The thermal loading, isolation, and retrieval requirements of a repository present unique technological challenges unknown to conventional mining practice. The technology issues and developments required in the areas of excavation, roof and ground support, equipment development, instrumentation development, and sealing are presented. Performance assessment methods must be developed to evaluate the adequacies of technologies developed to design, construct, operate, and decommission a repository. A stepwise test-and-development approach is used in the Basalt Waste Isolation Project to develop cost-effective technologies for a repository

  5. 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)

  6. Assessment of uncertainties in measurement of pH in hostile environments characteristic of nuclear repositories

    International Nuclear Information System (INIS)

    Kreider, K.G.; Tarlov, M.J.; Huang, P.H.

    1991-10-01

    This report focuses on evaluation and characteristics of sputtered thin film pH electrodes which can be used to assess the corrosivity of hot (100 degree C) aqueous solutions present in nuclear repositories. Sputtered thin films have the advantages of high temperature capability, ruggedness, and low cost. The iridium oxide films were found to have a linear, 58 mV/pH, response to changes in pH. They had little hysteresis but drifted approximately 0.2 V over a period of two days exposure to pH 2--12 solutions. The films were found to be insensitive to interference from most ions such as alkali ions but had redox sensitivity to ferri-/ferrocyanide ions. Although special surface treatments were needed for the films for good adherence at 200 degree C the films were not degraded after 20 hours exposure at pH 4, 7, and 10 at 200 degree C. Ruthenium oxide sputtered films performed equally well to the iridium oxide films in parallel tests. The report also contains information on electrochemistry and testing of thin film electrodes and the characterization of the thin films by x-ray photoemission spectroscopy, ultraviolet photoemission spectroscopy, and ion scattering spectroscopy. 123 refs., 29 figs., 7 tabs

  7. Geologic repositories for radioactive waste: the nuclear regulatory commission geologic comments on the environmental assessment

    International Nuclear Information System (INIS)

    Justus, P.S.; Trapp, J.S.; Westbrook, K.B.; Lee, R.; Blackford, M.B.; Rice, B.

    1985-01-01

    The NRC staff completed its review of the Environmental Assessments (EAs) issued by the Department of Energy (DOE) in December, 1984, in support of the site selection processes established by the Nuclear Waste Policy Act of 1982 (NWPA). The EAs contain geologic information on nine sites that DOE has identified as potentially acceptable for the first geologic repository in accordance with the requirements of NWPA. The media for the sites vary from basalt at Hanford, Washington, tuff at Yucca Mountain, Nevada, bedded salt in the Palo Duro Basin, Texas and Paradox Basin, Utah, to salt domes in Mississippi and Louisiana. Despite the diversity in media there are common areas of concern for all sites. These include; structural framework and pattern, rates of tectonic and seismic activity, characterization of subsurface features, and stratigraphic thickness, continuity and homogeneity. Site-specific geologic concerns include: potential volcanic and hydrothermal activity at Yucca Mountain, potential hydrocarbon targets and deep basalt and sub-basalt structure at Hanford, and potential dissolution at all salt sites. The NRC comments were influenced by the performance objectives and siting criteria of 10 CFR Part 60 and the environmental protection criteria in 40 CFR Part 191, the applicable standards proposed by EPA. In its review the NRC identified several areas of geologic concern that it recommended DOE re-examine to determine if alternative or modified conclusions are appropriate

  8. Tailored model abstraction in performance assessments

    International Nuclear Information System (INIS)

    Kessler, J.H.

    1995-01-01

    Total System Performance Assessments (TSPAs) are likely to be one of the most significant parts of making safety cases for the continued development and licensing of geologic repositories for the disposal of spent fuel and HLW. Thus, it is critical that the TSPA model capture the 'essence' of the physical processes relevant to demonstrating the appropriate regulation is met. But how much detail about the physical processes must be modeled and understood before there is enough confidence that the appropriate essence has been captured? In this summary the level of model abstraction that is required is discussed. Approaches for subsystem and total system performance analyses are outlined, and the role of best estimate models is examined. It is concluded that a conservative approach for repository performance, based on limited amount of field and laboratory data, can provide sufficient confidence for a regulatory decision

  9. Derivation of parameters necessary for the evaluation of performance of sites for deep geological repositories with particular reference to bedded salt, Livermore, California. Volume I. Main text

    International Nuclear Information System (INIS)

    Ashby, J.P.; Rawlings, G.E.; Soto, C.A.; Wood, D.F.; Chorley, D.W.

    1979-12-01

    A survey of parameters to be considered in the evaluation of sites for deep geologic nuclear waste repositories is presented. As yet, no comprehensive site selection procedure or performance evaluation approach has been adopted. A basis is provided for the development of parameters by discussing both site selection and performance evaluation. Three major groups of parameters are considered in this report: geologic, mining/rock mechanics, and hydrogeologic. For each type, the role of the parameter in the evaluation of repository sites is discussed. The derivation of the parameter by measurement, correlation, inference, or other method is discussed. Geologic parameters define the framework of the repository site and can be used in development of conceptual models and the prediction of long-term performance. Methods for deriving geological parameters include mapping, surveying, drilling, geophysical investigation, and historical and regional analysis. Rock mechanics/mining parameters are essential for the prediction of short-term performance and the development of initial conditions for modeling of long-term performance. Rock mechanics/mapping parameters can be derived by field or laboratory investigation, correlation, and theoretically or empirically based inference. Hydrogeologic parameters are the most important for assessment of long-term radionuclide confinement, since transport throughout the regional hydrogeologic system is the most likely mode of radionuclide escape from geologic repositories. Hydrogeologic parameters can be derived by hydrogeologic mapping and interpretation, hydrogeologic system modeling, field measurements, and lab tests. Procedures used in determination and statistical evaluation of geologic and rock mechanics parameters are discussed

  10. Derivation of parameters necessary for the evaluation of performance of sites for deep geological repositories with particular reference to bedded salt, Livermore, California. Volume I. Main text

    Energy Technology Data Exchange (ETDEWEB)

    Ashby, J.P.; Rawlings, G.E.; Soto, C.A.; Wood, D.F.; Chorley, D.W.

    1979-12-01

    A survey of parameters to be considered in the evaluation of sites for deep geologic nuclear waste repositories is presented. As yet, no comprehensive site selection procedure or performance evaluation approach has been adopted. A basis is provided for the development of parameters by discussing both site selection and performance evaluation. Three major groups of parameters are considered in this report: geologic, mining/rock mechanics, and hydrogeologic. For each type, the role of the parameter in the evaluation of repository sites is discussed. The derivation of the parameter by measurement, correlation, inference, or other method is discussed. Geologic parameters define the framework of the repository site and can be used in development of conceptual models and the prediction of long-term performance. Methods for deriving geological parameters include mapping, surveying, drilling, geophysical investigation, and historical and regional analysis. Rock mechanics/mining parameters are essential for the prediction of short-term performance and the development of initial conditions for modeling of long-term performance. Rock mechanics/mapping parameters can be derived by field or laboratory investigation, correlation, and theoretically or empirically based inference. Hydrogeologic parameters are the most important for assessment of long-term radionuclide confinement, since transport throughout the regional hydrogeologic system is the most likely mode of radionuclide escape from geologic repositories. Hydrogeologic parameters can be derived by hydrogeologic mapping and interpretation, hydrogeologic system modeling, field measurements, and lab tests. Procedures used in determination and statistical evaluation of geologic and rock mechanics parameters are discussed.

  11. Selection of nuclide decay chains for use in the assessment of the radiological impact of geological repositories for radioactive waste

    International Nuclear Information System (INIS)

    Thorne, M.C.

    1982-12-01

    The criteria for selecting nuclide decay chains for use in the assessment of the radiological impact of geological repositories for radioactive waste are given. The reduced chains recommended for use with SYVAC are described. (author)

  12. 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

  13. 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

  14. An assessment of gas impact on geological repository. Methodology and material property of gas migration analysis in engineered barrier system

    International Nuclear Information System (INIS)

    Yamamoto, Mikihiko; Mihara, Morihiro; Ooi, Takao

    2004-01-01

    Gas production in a geological repository has potential hazard, as overpressurisation and enhanced release of radionuclides. Amongst data needed for assessment of gas impact, gas migration properties of engineered barriers, focused on clayey and cementitious material, was evaluated in this report. Gas injection experiments of saturated bentonite sand mixture, mortar and cement paste were carried out. In the experiments, gas entry phenomenon and gas outflow rate were observed for these materials. Based on the experimental results, two-phase flow parameters were evaluated quantitatively. A conventional continuum two-phase flow model, which is only practically used multidimensional multi-phase flow model, was applied to fit the experimental results. The simulation results have been in good agreement with the gas entry time and the outflow flux of gas and water observed in the experiments. It was confirmed that application of the continuum two-phase flow model to gas migration in cementitious materials provides sufficient degree of accuracy for assessment of repository performance. But, for sand bentonite mixture, further extension of basic two-phase flow model is needed especially for effect of stress field. Furthermore, gas migration property of other barrier materials, including rocks, but long-term gas injection test, clarification of influence of chemicals environment and large-scale gas injection test is needed for multi-barrier assessment tool development and their verification. (author)

  15. Tools for Managing Repository Objects

    OpenAIRE

    Banker, Rajiv D.; Isakowitz, Tomas; Kauffman, Robert J.; Kumar, Rachna; Zweig, Dani

    1993-01-01

    working Paper Series: STERN IS-93-46 The past few years have seen the introduction of repository-based computer aided software engineering (CASE) tools which may finally enable us to develop software which is reliable and affordable. With the new tools come new challenges for management: Repository-based CASE changes software development to such an extent that traditional approaches to estimation, performance, and productivity assessment may no longer suffice - if they ever...

  16. Feasibility assessment of copper-base waste package container materials in a tuff repository

    International Nuclear Information System (INIS)

    Acton, C.F.; McCright, R.D.

    1986-01-01

    This report discussed progress made during the second year of a two-year study on the feasibility of using copper or a copper-base alloy as a container material for a waste package in a potential repository in tuff rock at the Yucca Mountain site in Nevada. Corrosion testing in potentially corrosive irradiated environments received emphasis during the feasibility study. Results of experiments to evaluate the effect of a radiation field on the uniform corrosion rate of the copper-base materials in repository-relevant aqueous environments are given as well as results of an electrochemical study of the copper-base materials in normal and concentrated J-13 water. Results of tests on the irradiation of J-13 water and on the subsequent formation of hydrogen peroxide are given. A theoretical study was initiated to predict the long-term corrosion behavior of copper in the repository. Tests were conducted to determine whether copper would adversely affect release rates of radionuclides to the environment because of degradation of the Zircaloy cladding. A manufacturing survey to determine the feasibility of producing copper containers utilizing existing equipment and processes was completed. The cost and availability of copper was also evaluated and predicted to the year 2000. Results of this feasibility assessment are summarized

  17. 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)

  18. 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

  19. Effects of silica redistribution on performance of high-level nuclear waste repositories in saturated geologic formations

    International Nuclear Information System (INIS)

    Verma, A.; Pruess, K.

    1985-11-01

    Evaluation of the thermohydrological conditions near high-level waste packages is needed for the design of the waste canister and for overall repository design and performance assessment. Most available studies in this area have assumed that the hydrologic properties of the host rock do not change in response to the thermal, mechanical or chemical effects caused by waste emplacement. However, the ramifications of this simplifying assumption have not been substantiated. We have studied dissolution and precipitation of silica in thermally driven flow systems, including changes in formation porosity and permeability. Using numerical simulation, we compare predictions of thermohydrological conditions with and without inclusion of silica redistribution effects. Two cases were studied, namely, a canister-scale problem, a repository-wide thermal convection problem, and different pore models were employed for the permeable medium (fractures with uniform or non-uniform cross sections). We find that silica redistribution generally has insignificant effects on host rock and canister temperatures, pore pressures, or flow velocites

  20. 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)

  1. Geochemical performance of earthen and cementitious sealing materials for radioactive waste repositories

    International Nuclear Information System (INIS)

    Melchoir, D.; Glazier, R.; Marton, R.

    1988-01-01

    Earthen and cementitious materials are proposed as part of the sealing system for radioactive waste repositories. Compacted clay-bearing earthen materials could be used in sealing shafts and shaft entryways; and in the waste emplacement boundary areas in some repository designs. Earthen material mixtures are being considered because they can be engineered and emplaced to achieve low permeabilities, appropriate swelling characteristics, and adequate strength with little tendency to degrade during changing environmental conditions. The proposed earthen sealing materials include sodium and calcium mont-morillonites, illites, and mixtures with graded aggregates of sand. To assess the relative advantages and disadvantages of various pure and mixed materials, important geochemical processes (e.g., ion-exchange, phase transformation, dissolution, and precipitation of secondary minerals) need to be evaluated. These processes could impact seal integrity by changing permeability and/or mineral swell potential. Hydrous calcium-silicate-based cementitious materials such as grouts or concrete might also be used in some proposed sealing systems

  2. Process of performance assessment

    International Nuclear Information System (INIS)

    King, C.M.; Halford, D.K.

    1987-01-01

    Performance assessment is the process used to evaluate the environmental consequences of disposal of radioactive waste in the biosphere. An introductory review of the subject is presented. Emphasis is placed on the process of performance assessment from the standpoint of defining the process. Performance assessment, from evolving experience at DOE sites, has short-term and long-term subprograms, the components of which are discussed. The role of mathematical modeling in performance assessment is addressed including the pros and cons of current approaches. Finally, the system/site/technology issues as the focal point of this symposium are reviewed

  3. Risk and uncertainty assessment for a potential HLW repository in Korea: TSPA 2006

    International Nuclear Information System (INIS)

    Hwang, Y.S.; Kang, C.H.

    2004-01-01

    KAERI has worked on the concept development on permanent disposal of HLW and its total system performance assessment since 1997. More than 36 000 MT of spent nuclear fuel from PWR and CANDU reactors is planned to be disposed of in crystalline bed-rocks. The total system performance assessment (TSPA) tools are under development. The KAERI FEP encyclopedia is actively developed to include all potential FEP suitable for Korean geo- and socio conditions. The FEPs are prioritized and then categorized to the intermediate level FEP groups. These groups become elements of the rock engineering system (RES) matrix. Then the sub-scenarios such as a container failure, groundwater migration, solute transport, etc are developed by connecting interactions between diagonal elements of the RES matrix. The full scenarios are developed from the combination of sub-scenarios. For each specific scenario, the assessment contexts and associated assessment method flow charts are developed. All information on these studies is recorded into the web based programme, FEAS (FEP to Assessment through Scenarios.) KAERI applies three basic programmes for the post closure radionuclide transport calculations; MASCOT-K, AMBER, and the new MDPSA under development. The MASCOT-K originally developed by Serco for a LLW repository has been extended extensively by KAERI to simulate release reactions such as congruent and gap releases in spent nuclear fuel. The new MDPSA code is dedicated for the probabilistic assessment of radio-nuclides in multi-dimensions of a fractured porous medium. To acquire input data for TSPA domestic experiment programmes as well as literature survey are performed. The data are stored in the Performance Assessment Input Data system (PAID.) To assure the transparency, traceability, retrievability, reproducibility, and review (T2R3) the web based KAERI QA system is developed. All tasks in TSPA are recorded under the concept of a 'Project' in this web system. Currently, FEAS, PAID

  4. Repository design

    Energy Technology Data Exchange (ETDEWEB)

    John, C M

    1982-01-01

    Various technical issues of radioactive waste design are addressed in this paper. Two approaches to repository design considered herein are: (1) design to minimize the disturbance of the hot rock; and (2) designs that intentionally modify the hot rock to insure better containment of the wastes. The latter designs range from construction of a highly impermeable barrier around a spherical cavern to creating a matrix of tunnels and boreholes to form a cage within which the hydraulic pressure is nearly constant. Examples of these design alternatives are described in some detail. It is concluded that proposed designs for repositories illustrate that performance criteria considered acceptable for such facilities can be met by appropriate site selection and repository engineering. With these technically feasible design concepts, it is also felt that socioeconomic and institutional issues can be better resolved. (BLM)

  5. Critical assessment of seismic and geomechanics literature related to a high-level nuclear waste underground repository

    Energy Technology Data Exchange (ETDEWEB)

    Kana, D.D.; Vanzant, B.W.; Nair, P.K. [Southwest Research Inst., San Antonio, TX (USA). Center for Nuclear Waste Regulatory Analyses; Brady, B.H.G. [ITASCA Consulting Group, Inc., Minneapolis, MN (USA)

    1991-06-01

    A comprehensive literature assessment has been conducted to determine the nature and scope of technical information available to characterize the seismic performance of an underground repository and associated facilities. Significant deficiencies were identified in current practices for prediction of seismic response of underground excavations in jointed rock. Conventional analytical methods are based on a continuum representation of the host rock mass. Field observations and laboratory experiments indicate that, in jointed rock, the behavior of the joints controls the overall performance of underground excavations. Further, under repetitive seismic loading, shear displacement develops progressively at block boundaries. Field observations correlating seismicity and groundwater conditions have provided significant information on hydrological response to seismic events. However, lack of a comprehensive model of geohydrological response to seismicity has limited the transportability conclusions from field observations. Based on the literature study, matters requiring further research in relation to the Yucca Mountain repository are identified. The report focuses on understanding seismic processes in fractured tuff, and provides a basis for work on the geohydrologic response of a seismically disturbed rock mass. 220 refs., 43 figs., 11 tabs.

  6. Critical assessment of seismic and geomechanics literature related to a high-level nuclear waste underground repository

    International Nuclear Information System (INIS)

    Kana, D.D.; Vanzant, B.W.; Nair, P.K.

    1991-06-01

    A comprehensive literature assessment has been conducted to determine the nature and scope of technical information available to characterize the seismic performance of an underground repository and associated facilities. Significant deficiencies were identified in current practices for prediction of seismic response of underground excavations in jointed rock. Conventional analytical methods are based on a continuum representation of the host rock mass. Field observations and laboratory experiments indicate that, in jointed rock, the behavior of the joints controls the overall performance of underground excavations. Further, under repetitive seismic loading, shear displacement develops progressively at block boundaries. Field observations correlating seismicity and groundwater conditions have provided significant information on hydrological response to seismic events. However, lack of a comprehensive model of geohydrological response to seismicity has limited the transportability conclusions from field observations. Based on the literature study, matters requiring further research in relation to the Yucca Mountain repository are identified. The report focuses on understanding seismic processes in fractured tuff, and provides a basis for work on the geohydrologic response of a seismically disturbed rock mass. 220 refs., 43 figs., 11 tabs

  7. Some viewpoints on reference biospheres in Finnish performance assessments

    International Nuclear Information System (INIS)

    Rasilainen, K.; Kattilakoski, E.; Suolanen, V.; Vieno, T.; Vuori, S.

    2002-01-01

    Viewpoints are presented concerning biosphere studies in performance assessments of nuclear waste disposal. The points are based on experiences from several Finnish performance assessments. The latest performance assessment for spent fuel disposal, TILA-99, was considered in the Decision in Principle process for the site selection of the repository. The points given are also based on experiences from participation in international projects dealing with biosphere modelling, for instance BIOMOVS and BIOMASS. (author)

  8. Assessing microbiologically induced corrosion of waste package materials in the Yucca Mountain repository

    Energy Technology Data Exchange (ETDEWEB)

    Horn, J. M., LLNL

    1998-01-01

    The contribution of bacterial activities to corrosion of nuclear waste package materials must be determined to predict the adequacy of containment for a potential nuclear waste repository at Yucca Mountain (YM), NV. The program to evaluate potential microbially induced corrosion (MIC) of candidate waste container materials includes characterization of bacteria in the post-construction YM environment, determination of their required growth conditions and growth rates, quantitative assessment of the biochemical contribution to metal corrosion, and evaluation of overall MIC rates on candidate waste package materials.

  9. Nuclear waste repository design and construction

    International Nuclear Information System (INIS)

    Bohlke, B.M.; Monsees, J.E.

    1987-01-01

    Extensive underground excavation will be required for construction of a mined geologic repository for nuclear waste. Hundreds of thousands of feet of drift will be required based on the conceptual layout design for each candidate nuclear waste repository. Comparison of boring and blasting excavation methods are discussed, as are special design and construction requirements (e.g., quality assurance procedures and performance assessment) for the nuclear waste repository. Comparisons are made between boring and blasting construction methods for the repository designs proposed for salt, volcanic tuff, and basalt

  10. Repository exploration

    International Nuclear Information System (INIS)

    Pentz, D.L.

    1984-01-01

    This paper discusses exploration objectives and requirements for a nuclear repository in the U.S.A. The importance of designing the exploration program to meet the system performance objectives is emphasized and some examples of the extent of exploration required before the License Application for Construction Authorization is granted are also discussed

  11. 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).

  12. Spent fuel repositories and environmental impact assessment; the legal situation i Finland

    International Nuclear Information System (INIS)

    Manninen, J.

    1995-01-01

    The Finnish Act on Environmental Impact Assessment (468/94) builds on the EU Directive 85/337 on the assessment of the effects of certain public and private projects on the environment. The Act is supplemented by the Decree on Environmental Impact Assessment (792/88). The EIA Act is applicable to all nuclear facilities, spent fuel repositories included. In practice the new Act does not introduce any drastic changes to the legal situation, as far as nuclear facilities are concerned. The Finnish Nuclear Energy Act (990/87), which entered into force 1988, already contains most of the basic elements of the environmental impact assessment. In the article, the different steps in the Finnish EIA procedure is described

  13. Preliminary assessment of nuclear waste transportation cost and risk for operation of the first repository at candidate sites

    International Nuclear Information System (INIS)

    Peterson, R.W.; McSweeney, T.I.; Varadarajan, R.V.; Wilmot, E.L.; Cashwell, J.W.; Joy, D.S.

    1983-01-01

    To support the selection of the first commercial nuclear waste repository site in 1987, environmental analyses of five candidate site locations are currently being performed. The five locations are in the Gulf Interior Region, the Permian Basin, the Paradox Basin, Yucca Mountain and the Hanford reservation. Costs and operational risks associated with the transportation of nuclear wastes to a single repository located in these regions have been calculated for a life-cycle of 26 years

  14. WIPP performance assessment: impacts of human intrusion

    International Nuclear Information System (INIS)

    Anderson, D.R.; Hunter, R.L.; Bertram-Howery, S.G.; Lappin, A.R.

    1989-01-01

    The Waste Isolation Pilot Plant (WIPP) in southeastern New Mexico is a research and development facility that may become the USA's first and only mined geologic repository for transuranic waste. Human intrusion into the WIPP repository after closure has been shown by preliminary sensitivity analyses and calculations of consequences to be an important, and perhaps the most important, factor in long-term repository performance

  15. 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

  16. NRC performance assessment program

    International Nuclear Information System (INIS)

    Coplan, S.M.

    1986-01-01

    The U.S. Nuclear Regulatory Commission's (NRC) performance assessment program includes the development of guidance to the U.S. Department of Energy (DOE) on preparation of a license application and on conducting the studies to support a license application. The nature of the licensing requirements of 10 CFR Part 60 create a need for performance assessments by the DOE. The NRC and DOE staffs each have specific roles in assuring the adequacy of those assessments. Performance allocation is an approach for determining what testing and analysis will be needed during site characterization to assure that an adequate data base is available to support the necessary performance assessments. From the standpoint of establishing is implementable methodology, the most challenging performance assessment needed for licensing is the one that will be used to determine compliance with the U.S. Environmental Protection Agency's (EPA) containment requirement

  17. The impact of episodic nonequilibrium fracture-matrix flow on geological repository performance

    International Nuclear Information System (INIS)

    Buscheck, T.A.; Nitao, J.J.; Chestnut, D.A.

    1991-01-01

    Adequate representation of fracture-matrix interaction during episodic infiltration events is crucial in making valid hydrological predictions of repository performance at Yucca Mountain. Various approximations have been applied to represent fracture-matrix flow interaction, including the Equivalent Continuum Model (ECM), which assumes capillary equilibrium between fractures and matrix, and the Fracture-Matrix Model (FMM), which accounts for nonequilibrium fracture-matrix flow. We analyze the relative impact of matrix imbibition on episodic nonequilibrium fracture-matrix flow for the eight major hydrostratigraphic units in the unsaturated zone at Yucca Mountain. Comparisons are made between ECM and FMM predictions to determine the applicability of the ECM. The implications of nonequilibrium fracture-matrix flow on radionuclide transport are also discussed

  18. Assessment and Measurements of Degradation Processes in the Engineering Barriers of LILW Repository

    International Nuclear Information System (INIS)

    Zeleznik, N.; Sajna, A.; Petkovsek, B.

    2013-01-01

    In 2009 the Slovenian national agency for radwaste management (ARAO) adopted the national spatial plan for the low and intermediate level waste (LILW) repository to be located in Krsko municipality near NPP Krsko. The selected option for disposal was based on a silo type structure for the near surface disposal facility that will be situated close to a saturated aquifer. The soil in the region can be described as silt that extends a few hundred meters deep. The silt also contains sections of sand or clay. As the possibility exists that the natural geological barrier system will not be able to contain radionuclide migration it is a pre requisite that the proposed LILW repository must install engineered barriers. Research on different cementation materials are currently underway in order to find sustainable materials for the manufacturing of engineered barriers for the repository (silo, backfilling, concrete containers). The research also includes the assessment of possible site specific degradation processes in order to provide a methodology for the selection of appropriate locally available materials that will minimize the degradation processes. The research methodology was based on studying the characteristics (workability, compressive strength), durability (resistance to penetration of water, freeze/thaw resistance, resistance to groundwater), rheology (heat of hydration, autogenous and concrete shrinkage) and reinforcement corrosion of different concrete compositions. The characterization results were used to develop a numerical model for degradation processes to be found in the current concrete compositions. Although initial results indicated that the current concrete compositions are satisfactory, the research must be extended to the addition of binder materials to improve the characteristics of the manufactures concrete before degradation processes can be studied. (author)

  19. Procedures for uncertainty and sensitivity analysis in repository performance assessment

    International Nuclear Information System (INIS)

    Poern, K.; Aakerlund, O.

    1985-10-01

    The objective of the project was mainly a literature study of available methods for the treatment of parameter uncertainty propagation and sensitivity aspects in complete models such as those concerning geologic disposal of radioactive waste. The study, which has run parallel with the development of a code package (PROPER) for computer assisted analysis of function, also aims at the choice of accurate, cost-affective methods for uncertainty and sensitivity analysis. Such a choice depends on several factors like the number of input parameters, the capacity of the model and the computer reresources required to use the model. Two basic approaches are addressed in the report. In one of these the model of interest is directly simulated by an efficient sampling technique to generate an output distribution. Applying the other basic method the model is replaced by an approximating analytical response surface, which is then used in the sampling phase or in moment matching to generate the output distribution. Both approaches are illustrated by simple examples in the report. (author)

  20. Radionuclide transport modelling. Performance assessment of repositories in clays

    International Nuclear Information System (INIS)

    Ruebel, A.; Becker, D.A.; Fein, E.

    2007-12-01

    This study aims at the structural characterisation of embedded flame retardants and polymers. Analysis includes determination of the phase contents as well as description of their mutual interactions as a function of their thermal history in different atmospheres (air and nitrogen). Solid-state NMR spectroscopy is the most important tool applied in this thesis. This technique enables structural information about the amorphous state of the decomposed products, determination of the thermal and thermo-oxidative conversion in different amorphous and crystalline structures, and predictions about the short-range order of the observed nuclei. In this context measurements of single pulse (SP), rotor-synchronised spin echo (RSE) and cross polarisation (CP) experiments, as well as REDOR (Rotational echo double resonance) and TRAPDOR-(Transfer of population in double resonance) have been conducted. Additionally, 11 B SP and 31 P RSE experiments have been used to study quantitatively the borate and both phosphorus and phosphate contents in the solid residues. Particular emphasis was put on quantifying crystalline and non-crystalline phosphate phases. The first system consists of a combination of the flame retardants red phosphorus (P red ) and Mg(OH) 2 in HIPS. From the results of extensive NMR experiments it was deduced that most of inserted red phosphorus remains in amorphous phosphates phases (ortho, pyro, and chain/ring phosphates) besides some crystalline phosphate phases. Decomposition characteristics such as temperature and the availability of oxygen (N 2 atmosphere/air) show a significant influence on the decomposition process and the formation of phosphate phases (crystalline/amorphous). By comparing the results of the annealing processes to the results of the cone calorimeter measurements an anaerobic decomposition way was confirmed. In a second system, thermal reactions between the two flame retardants BDP and zinc borate as well their influence on the thermal decomposition of a PC/ABS blend were investigated. The thermally residues are independently of the atmosphere dominated by amorphous phosphate units. During the thermal decomposition the formation of α-Zn 3 (PO 4 ) 2 and BPO 4 could be identified as solid reaction products of the inserted flame retardants. In addition, the α-Zn 3 (PO 4 ) 2 /BPO 4 ratio served as an indicator of aerobic or anaerobic decomposition processes, proving an anaerobic decomposition mechanism for the fire residues

  1. Studies of colloids and their importance for repository performance assessment

    International Nuclear Information System (INIS)

    Laaksoharju, M.; Skaarman, C.; Degueldre, C.

    1995-12-01

    The processes, parameters and data used to evaluate the potential of nuclide transport by a colloid facilitated mechanism are reviewed and discussed in this report. Both steady-state (present situation) and possible future non-steady-state hydrogeochemistry in the geosphere are covered. In the steady-state scenario, the colloid (clay, silica, iron(III)hydroxide) concentration is around 20-45 micrograms/l which is considered to be a low value. The low colloid concentration is justified by the large attachment factor to the rock which reduces the stability of the colloids in the aquifer. Both reversible and irreversible sorption processes are reviewed. In the non-steady-state scenario, changes of hydrogeochemical properties may induce larger colloid concentrations. The increase of concentration is however limited and relaxation is always observed after any change. Emphasis is placed on the glaciation-deglaciation scenario. 53 refs, 12 figs, 3 tabs

  2. Assessing the performance of the Nagra HLW disposal concept

    International Nuclear Information System (INIS)

    Smith, P.; Zuidema, P.; McKinley, I.G.

    1995-01-01

    This article outlines the procedures used in safety assessment and illustrates their application in evaluating the performance of a high-level waste repository. Nagra's general safety assessment methodology has five main components: formulating the aims of the analysis, defining the safety concept, scenario development, consequence analysis and interpretation of results. A safety analysis based on conservative assumptions shows that the engineered barriers of the high-level waste repository are very effective in preventing release of radionuclides; this alone is sufficient to ensure that regulatory requirements can be met. The function of the host rock is to provide a favourable environment for the engineered barrier system. (author) 8 figs

  3. 10 CFR Appendix I to Part 960 - NRC and EPA Requirements for Postclosure Repository Performance

    Science.gov (United States)

    2010-01-01

    ... SCREENING OF POTENTIAL SITES FOR A NUCLEAR WASTE REPOSITORY Pt. 960, App. I Appendix I to Part 960—NRC and... after disposal (a) releases of radioactive materials to the accessible environment that are estimated to...,000 years for a repository containing wastes generated from 100,000 metric tons of heavy metal of...

  4. A comparative Study between GoldSim and AMBER Based Biosphere Assessment Models for an HLW Repository

    International Nuclear Information System (INIS)

    Lee, Youn-Myoung; Hwang, Yong-Soo

    2007-01-01

    To demonstrate the performance of a repository, the dose exposure rate to human being due to long-term nuclide releases from a high-level waste repository (HLW) should be evaluated and the results compared to the dose limit presented by the regulatory bodies. To evaluate such a dose rate to an individual, biosphere assessment models have been developed and implemented for a practical calculation with the aid of such commercial tools as AMBER and GoldSim, both of which are capable of probabilistic and deterministic calculation. AMBER is a general purpose compartment modeling tool and GoldSim is another multipurpose simulation tool for dynamically modeling complex systems, supporting a higher graphical user interface than AMBER and a postprocessing feature. And also unlike AMBER, any kind of compartment scheme can be rather simply constructed with an appropriate transition rate between compartments, GoldSim is designed to facilitate the object-oriented modules to address any specialized programs, similar to solving jig saw puzzles. During the last couple of years a compartment modeling approach for a biosphere has been mainly carried out with AMBER in KAERI in order to conservatively or rather roughly provide dose conversion factors to get the final exposure rate due to a nuclide flux into biosphere over various geosphere-biosphere interfaces (GBIs) calculated through nuclide transport modules. This caused a necessity for a newly devised biosphere model that could be coupled to a nuclide transport model with less conservatism in the frame of the development of a total system performance assessment modeling tool, which could be successfully done with the aid of GoldSim. Therefore, through the current study, some comparison results of the AMBER and the GoldSim approaches for the same case of a biosphere modeling without any consideration of geosphere transport are introduced by extending a previous study

  5. Final Systems Development Report for the Clark County Socioeconomic Impact Assessment of the Proposed High-Level Nuclear Waste Repository at Yucca Mountain, NV

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1992-06-18

    The Systems Development Report represents the third major step in the Clark County Socioeconomic Impact Assessment of the Proposed High-Level Nuclear Waste Repository at Yucca Mound Nevada. The first of these steps was to forge a Research Design that would serve as a guide for the overall research process. The second step was the construction of the Base Case, the purpose of which was to describe existing conditions in Clark County in the specified analytic areas of Economic-Demographic/Fiscal, Emergency Planning and Management, Transportation and Sociocultural analysis. The base case description will serve as a basis for assessing changes in these topic areas that might result from the Yucca Mountain project. These changes will be assessed by analyzing conditions with and without repository development in the county. Prior to performing such assessments, however, the snapshot type of data found in the base case must be operationalized or systematized to allow for more dynamic data utilization. In other words, a data system that can be used to analyze the consequences of the introduction of different variables (or variable values) in the Clark County context must be constructed. Such a system must be capable of being updated through subsequent data collection and monitoring efforts to both provide a rolling base case and supply information necessary to construct trend analyses. For example, during the Impact Assessment phase of the study process, the without repository analysis is accomplished by analyzing growth for the county given existing conditions and likely trends. These data are then compared to the with Yucca Mountain project conditions anticipated for the county. Similarly, once the emergency planning management and response needs associated with the repository are described, these needs will be juxtaposed against existing (and various future) capacity(ies) in order to determine the nature and magnitude of impacts in this analytic area. Analogous tasks

  6. 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

  7. 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

  8. Environmental assessment for OPG's deep geologic repository for low and intermediate level waste

    International Nuclear Information System (INIS)

    Barker, D.; Rawlings, M.; Beal, A.

    2011-01-01

    The environmental assessment process for the Deep Geologic Repository (DGR) Project was initiated very early in the planning stages. Feasibility studies were initiated in 2003, after Ontario Power Generation (OPG) and the Municipality of Kincardine signed a Memorandum of Understanding agreeing to assess options for long-term management of low and intermediate level waste (L and ILW) options at the Bruce nuclear site. The location of the DGR, in the Municipality of Kincardine, is based on a willing and informed host community. The preferred approach, the DGR at the Bruce nuclear site, was advanced based on results of feasibility studies which looked at a number of options for long-term management of L&ILW and support from the local community and their elected representatives. The federal environmental assessment of the project was initiated following the signing of a Host Community Agreement and completion of a telephone poll, the results of which indicated that the majority of Municipality of Kincardine residents support the project. The environmental assessment began in 2006 as a comprehensive study and was ultimately referred to a joint review panel process in 2009. The environmental assessment considers the potential near-term effects of the construction and operations of the proposed project. Because of the nature of the project, the assessment of effects also considers long-term effects extending out to the million year time-frame, including effects of climate change, glaciations and seismic activity. (author)

  9. Performance assessment of DOE spent nuclear fuel and surplus plutonium

    International Nuclear Information System (INIS)

    Duguid, J.O.; Vallikat, V.; McNeish, J.

    1998-01-01

    Yucca Mountain, in southern Nevada, is under consideration by the US Department of Energy (DOE) as a potential site for the disposal of the nation's radioactive wastes in a geologic repository. The wastes consist of commercial spent fuel, DOE spent nuclear fuel (SNF), high level waste (HLW), and surplus plutonium. The DOE was mandated by Congress in the fiscal 1997 Energy and Water Appropriations Act to complete a viability assessment (VA) of the repository in September of 1998. The assessment consists of a preliminary design concept for the critical elements of the repository, a total system performance assessment (TSPA), a plan and cost estimate for completion of the license application, and an estimate of the cost to construct and operate the repository. This paper presents the results of the sensitivity analyses that were conducted to examine the behavior of DOE SNF and plutonium waste forms in the environment of the base case repository that was modeled for the TSPA-VA. Fifteen categories of DOE SNF and two Plutonium waste forms were examined and their contribution to radiation dose to humans was evaluated

  10. U.S. Department of Energy approaches to the assessment of radionuclide migration for the geologic repository program

    International Nuclear Information System (INIS)

    Luik, A.E. van; Apted, M.J.

    1988-01-01

    Potential radionuclide migration in geologic repositories is being addressed by the U.S. Department of Energy's Office of Civilian Radioactive Waste Management through its Office of Geologic Repositories (OGR). A diversity of geohydrologic settings is being investigated: unsaturated tuff, saturated basalt, and bedded salt. A number of approaches to assessing potential migration are being considered. Mass transfer is prominent among near-field approaches. For far-field analysis of migration in the geosphere, detailed characterizations of potential repository sites will lead to site-specific models describing radionuclide migration for a variety of postulated release scenarios. Finite-element and finite-difference codes are being used and developed to solve the mathematical equations pertinent to far-field assessments. Computational approaches presently in use generally require distribution coefficients to estimate the retardation of specific radionuclides with respect to the transport rate of water. 26 refs

  11. Review on assessment methodology for human intrusion into a repository for radioactive waste

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Dong Keun; Kim, Jung Woo; Jeong, Jong Tae; Baik, Min Hoon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-09-15

    An approach to assess inadvertent human intrusion into radwaste repository was proposed with the assumption that the intrusion occurs after a loss of knowledge of the hazardous nature of the disposal facility. The essential boundary conditions were derived on the basis of international recommendations, followed by an overall approach to deal with inadvertent human intrusion. The interrelation between societal factors, human intrusion scenarios, and protective measures is described to provide a concrete explanation of the approach, including the detailed procedures to set up the human intrusion scenario. The procedure for deriving protective measures is also explained with four steps, including how to derive a safety framework, general measures, potential measures, and eventual protective measures on the basis of stylized scenarios. It is expected that the approach proposed in this study will be used effectively to reduce the potential for and/or the consequences of human intrusion during the entire process of realizing a disposal facility.

  12. Review on assessment methodology for human intrusion into a repository for radioactive waste

    International Nuclear Information System (INIS)

    Cho, Dong Keun; Kim, Jung Woo; Jeong, Jong Tae; Baik, Min Hoon

    2016-01-01

    An approach to assess inadvertent human intrusion into radwaste repository was proposed with the assumption that the intrusion occurs after a loss of knowledge of the hazardous nature of the disposal facility. The essential boundary conditions were derived on the basis of international recommendations, followed by an overall approach to deal with inadvertent human intrusion. The interrelation between societal factors, human intrusion scenarios, and protective measures is described to provide a concrete explanation of the approach, including the detailed procedures to set up the human intrusion scenario. The procedure for deriving protective measures is also explained with four steps, including how to derive a safety framework, general measures, potential measures, and eventual protective measures on the basis of stylized scenarios. It is expected that the approach proposed in this study will be used effectively to reduce the potential for and/or the consequences of human intrusion during the entire process of realizing a disposal facility

  13. 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.

  14. 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

  15. Effects of heat from high-level waste on performance of deep geological repository components

    International Nuclear Information System (INIS)

    1984-11-01

    This report discusses the effects of heat on the deep geological repository systems and its different components. The report is focussed specifically on effects due to thermal energy release solely from high-level waste or spent fuel. It reviews the experimental data and theoretical models of the effects of heat both on the behaviour of engineered and natural barriers. A summary of the current status of research and repository development including underground test facilities is presented

  16. Assessing Scientific Performance.

    Science.gov (United States)

    Weiner, John M.; And Others

    1984-01-01

    A method for assessing scientific performance based on relationships displayed numerically in published documents is proposed and illustrated using published documents in pediatric oncology for the period 1979-1982. Contributions of a major clinical investigations group, the Childrens Cancer Study Group, are analyzed. Twenty-nine references are…

  17. Context for performance assessment

    International Nuclear Information System (INIS)

    Kocher, D.C.

    1997-01-01

    In developing its recommendations on performance assessment for disposal of low-level radioactive waste, Scientific committee 87-3 of the National Council on Radiation Protection and Measurements (NCRP) has considered a number of topics that provide a context for the development of suitable approaches to performance assessment. This paper summarizes the Committee' discussions on these topics, including (1) the definition of low-level waste and its sources and properties, as they affect the variety of wastes that must be considered, (2) fundamental objectives and principles of radioactive waste disposal and their application to low-level waste, (3) current performance objectives for low-level waste disposal in the US, with particular emphasis on such unresolved issues of importance to performance assessment as the time frame for compliance, requirements for protection of groundwater and surface water, inclusion of doses from radon, demonstrating compliance with fixed performance objectives using highly uncertain model projections, and application of the principle that releases to the environment should be maintained as low as reasonably achievable (ALARA), (4) the role of active and passive institutional controls over disposal sites, (5) the role of the inadvertent human intruder in low-level waste disposal, (6) model validation and confidence in model outcomes, and (7) the concept of reasonable assurance of compliance

  18. Texas' performance assessment work

    International Nuclear Information System (INIS)

    Charbeneau, R.J.; Hertel, N.E.; Pollard, C.G.

    1990-01-01

    The Texas Low-Level Radioactive Waste Disposal Authority is completing two years of detailed on-site suitability studies of a potential low-level radioactive waste disposal site in Hudspeth County, Texas. The data from these studies have been used to estimate site specific parameters needed to do a performance assessment of the site. The radiological impacts of the site have been analyzed as required for a license application. The approach adopted for the performance assessment was to use simplified and yet conservative assumptions with regard to releases, radionuclide transport, and dose calculations. The methodologies employed in the performance assessment are reviewed in the paper. Rather than rely on a single computer code, a modular approach to the performance assessment was selected. The HELP code was used to calculate the infiltration rate through the trench covers and the amount of leachate released from this arid site. Individual pathway analyses used spreadsheet calculations. These calculations were compared with those from other computer models including CRRIS, INGDOS, PATHRAE, and MICROSHIELD copyright, and found to yield conservative estimates of the effective whole body dose. The greatest difficulty in performing the radiological assessment of the site was the selection of reasonable source terms for release into the environment. A surface water pathway is unreasonable for the site. Though also unlikely, the groundwater pathway with exposure through a site boundary well was found to yield the largest calculated dose. The more likely pathway including transport of leachate from the facility through the unsaturated zone and returning to the ground surface yields small doses. All calculated doses associated with normal releases of radioactivity are below the regulatory limits

  19. Performance of phosphoric acid activated montmorillonite as buffer materials for radioactive waste repository

    International Nuclear Information System (INIS)

    Wang, Tsing-Hai; Liu, Tsung-Ying; Wu, Ding-Chiang; Li, Ming-Hsu; Chen, Jiann-Ruey; Teng, Shi-Ping

    2010-01-01

    In this study, the performance of phosphoric acid activated montmorillonite (PAmmt) was evaluated by cesium ions adsorption experiments. The PAmmt samples were obtained by activating with 1, 3 and 5 mol L -1 of phosphoric acid, respectively under reflux for 3, 12, and 24 h. Experimental results demonstrated that the treatment of raw K-10 montmorillonite with phosphoric acid increased the materials' affinity for Cs uptake and no significant amount of suspension solids were produced. A relatively insignificant variation in the CEC value was observed. Furthermore, PAmmt also showed high adsorption selectivity toward Cs ions. The improved sorptive properties were mainly related to the increased surface area and the relatively higher surface charge density. Increased specific surface area was the resulted from partial decomposition of lamellar structure of mmt; while the higher surface charge density was caused by the protonation of octahedral Al-OH sites during the acid activation. Generally speaking, stronger acid concentration and longer activation times would produce relatively more decomposed PAmmt particles. However, as the activation exceeds 3 h, the precipitation of Si 4+ would passivate PAmmt against further acid attacks. Based upon our results, acid activation by phosphoric acid could produce PAmmt samples with high sorption capacity and selectivity, and good structural integrity, which are beneficial to be used at radioactive waste repository.

  20. Performance assessment of nuclear waste isolation systems

    International Nuclear Information System (INIS)

    Lee, W.L.

    1984-01-01

    A number of concepts have been proposed for the isolation of highly radioactive wastes, and it will be necessary to demonstrate the safety of such systems. In many countries including the U.S., the waste isolation system of choice is deep mined geologic repositories. Because of the complex nature of the multiple isolation barriers afforded by mined geologic disposal systems, and the long isolation periods involved, this demonstration can only be indirect. In recent years this indirect demonstration, mostly through mathematical modeling, is called performance assessment. Performance Assessment can be defined to mean the development, testing, and application of a series of mathematical models and computer codes which traces the movement of radionuclides from a waste isolation system to the biosphere and any resultant dose to man. In modeling such a repository system, it is often convenient to divide it into a number of subsystems, there may be several different processes that need to be modeled, individually and interactively. For instance, this waste package will probably consist of a waste form such as borosilicate glass containing the radioisotopes, a canister, an overpack material such as steel or copper, and a buffer material such as bentonite. The processes to be modeled at the waste package scale include radioisotope inventory and decay, thermal radiation, radiolysis effects, corrosion, leading and fluid flow. In tracing radionuclide transport through rock, the processes of importance are probably groundwater flow, and sorption and retardation of radionuclide movement

  1. Putting HLW performance assessment results in perspective

    International Nuclear Information System (INIS)

    Neall, F.; Smith, P.; Sumerling, T.; Umeki, H.

    1995-01-01

    According to performance assessment results for the different disposal concepts investigated, the maximum radiation doses to the population lie well below the limit set in the official Swiss Protection Objective and below the level of present-day natural background radiation. A comparison of different performance assessments has shown that the following key factors determine radionuclide release from a repository: radionuclide inventory, canister material and failure mode, nuclide solubility limits, the permeability of the buffer material, retardation during transport through the near-field, the presence of an excavation disturbed zone in the rock, the distance to the nearest major water-bearing fracture zone, the conceptual model for transport in fractured rock and near-surface dilution and dose factors. (author) 2 figs., 2 tabs

  2. 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

  3. Principal elements of the Basalt Waste Isolation Project performance assessment studies

    International Nuclear Information System (INIS)

    Baca, R.G.; Wilde, R.T.

    1983-12-01

    In this paper, three of the principal elements of the Basalt Waste Isolation Project (BWIP) performance assessment studies are focused on: (1) development of a methodology for probabilistic risk assessment, (2) performance analyses of repository subsystems, and (3) selection and ranking of disruptive event scenarios. Other elements of the BWIP performance assessment studies are briefly outlined. 12 references, 5 figures

  4. AEGIS technology demonstration for a nuclear waste repository in basalt. Assessment of effectiveness of geologic isolation systems

    International Nuclear Information System (INIS)

    Dove, F.H.; Cole, C.R.; Foley, M.G.

    1982-09-01

    A technology demonstration of current performance assessment techniques as applied to a nuclear waste repository in the Columbia Plateau Basalts was conducted. Hypothetical repository coordinates were selected for an actual geographical setting on the Hanford Reservation in the state of Washington. Published hydrologic and geologic data used in the analyses were gathered in 1979 or earlier. The following report documents the technology demonstration in basalt. Available information has been used to establish the data base and initial hydrologic and geologic interpretations for this site-specific application. A simplified diagram of the AEGIS analyses is shown. Because an understanding of the dynamics of ground-water flow is essential to the development of release scenarios and consequence analyses, a key step in the demonstration is the systems characterization contained in the conceptual model. Regional and local ground-water movement patterns have been defined with the aid of hydrologic computer models. Hypothetical release scenarios have been developed and evaluated by a process involving expert opinion and a Geologic Simulation Model for basalt. (The Geologic Simulation Model can also be used to forecast future boundary conditions for the hydrologic simulation.) Chemical reactivity of the basalt with ground water will influence the leaching and transport of radionuclides; solubility equilibria based on available data are estimated with geochemical models. After the radionuclide concentrations are mathematically introduced into the ground-water movement patterns, waste movement patterns are outlined over elapsed time. Contaminant transport results are summarized for significant radionuclides that are hypothetically released to the accessible environment and to the biosphere

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

    KAUST Repository

    Falth, B.

    2014-12-09

    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 is surrounded by a number of smaller discontinuities representing fractures on which secondary displacements may be induced. Initial stresses are applied and a rupture is initiated at a predefined hypocenter and propagated at a specified rupture speed. During rupture we monitor shear displacements taking place on the nearby fracture planes in response to static as well as dynamic effects. As a numerical tool, we use the 3Dimensional Distinct Element Code (3DEC) because it has the capability to handle numerous discontinuities with different orientations and at different locations simultaneously. In tests performed to benchmark the capability of our method to generate and propagate seismic waves, 3DEC generates results in good agreement with results from both Stokes solution and the Compsyn code package. In a preliminary application of our method to the nuclear waste repository site at Forsmark, southern Sweden, we assume end-glacial stress conditions and rupture on a shallow, gently dipping, highly prestressed fault with low residual strength. The rupture generates nearly complete stress drop and an M-w 5.6 event on the 12 km(2) rupture area. Of the 1584 secondary fractures (150 m radius), with a wide range of orientations and locations relative to the fault, a majority move less than 5 mm. The maximum shear displacement is some tens of millimeters at 200 m fault-fracture distance.

  6. AEGIS technology demonstration for a nuclear waste repository in basalt. Assessment of effectiveness of geologic isolation systems

    Energy Technology Data Exchange (ETDEWEB)

    Dove, F.H.; Cole, C.R.; Foley, M.G.

    1982-09-01

    A technology demonstration of current performance assessment techniques as applied to a nuclear waste repository in the Columbia Plateau Basalts was conducted. Hypothetical repository coordinates were selected for an actual geographical setting on the Hanford Reservation in the state of Washington. Published hydrologic and geologic data used in the analyses were gathered in 1979 or earlier. The following report documents the technology demonstration in basalt. Available information has been used to establish the data base and initial hydrologic and geologic interpretations for this site-specific application. A simplified diagram of the AEGIS analyses is shown. Because an understanding of the dynamics of ground-water flow is essential to the development of release scenarios and consequence analyses, a key step in the demonstration is the systems characterization contained in the conceptual model. Regional and local ground-water movement patterns have been defined with the aid of hydrologic computer models. Hypothetical release scenarios have been developed and evaluated by a process involving expert opinion and a Geologic Simulation Model for basalt. (The Geologic Simulation Model can also be used to forecast future boundary conditions for the hydrologic simulation.) Chemical reactivity of the basalt with ground water will influence the leaching and transport of radionuclides; solubility equilibria based on available data are estimated with geochemical models. After the radionuclide concentrations are mathematically introduced into the ground-water movement patterns, waste movement patterns are outlined over elapsed time. Contaminant transport results are summarized for significant radionuclides that are hypothetically released to the accessible environment and to the biosphere.

  7. 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

  8. Contents of a regulatory strategy for assessing future human actions in the safety evaluation of a repository for spent fuels

    International Nuclear Information System (INIS)

    Wilmot, R.D.; Wickham, S.M.; Galson, D.A.

    2001-08-01

    The objective of this report is to discuss issues that should be considered in the development of a regulatory strategy for assessing future human actions in any forthcoming license application for a deep repository for spent fuel in Sweden and for sites of other repositories. The report comprises an outline of key issues concerning the treatment of future human actions in safety assessment, reviews of regulatory developments, recent safety assessments and supporting studies, and international initiatives on the treatment of future human actions in safety assessment, and the principal elements of a regulatory strategy. Performance assessments (PAs) are generally accepted as providing illustrations of system performance under given sets of assumptions. The results of PAs are clearer and easier to understand if certain large uncertainties are accounted for by determining performance under several different sets of assumptions or scenarios, each of which defines a possible evolution of the disposal system. A number of assumptions can be made that would restrict the scope of an assessment without reducing the credibility of the corresponding safety case. Reducing speculation about technological development, by assuming that the techniques used in future human activities are similar to those currently in use in the region or at similar sites, will simplify the assessment. A distinction is generally made between inadvertent and intentional intrusion, with intentional activities excluded because society cannot protect future populations from their own actions if they understand the potential consequences. A division of human activities into 'recent and ongoing' and 'future' activities considers not only the timing of the activities but also the degree of control or influence that can be imposed on them. Recent and ongoing human activities are those that affect an area beyond the immediate vicinity of the disposal facility and which neither the proponent nor the regulator

  9. Energy performance assessment methodology

    Energy Technology Data Exchange (ETDEWEB)

    Platzer, W.J. [Fraunhofer Inst. for Solar Energy Systems, Freiburg (Germany)

    2006-01-15

    The energy performance of buildings are intimately connected to the energy performance of building envelopes. The better we understand the relation between the quality of the envelope and the energy consumption of the building, the better we can improve both. We have to consider not only heating but all service energies related to the human comfort in the building, such as cooling, ventilation, lighting as well. The complexity coming from this embracing approach is not to be underestimated. It is less and less possible to realted simple characteristic performance indicators of building envelopes (such as the U-value) to the overall energy performance. On the one hand much more paramters (e.g. light transmittance) come into the picture we have to assess the product quality in a multidimensional world. Secondly buildings more and more have to work on a narrow optimum: For an old, badly insulated building all solar gains are useful for a high-performance building with very good insulation and heat recovery systems in the ventilation overheating becomes more likely. Thus we have to control the solar gains, and sometimes we need high gains, sometimes low ones. And thirdly we see that the technology within the building and the user patterns and interactions as well influence the performance of a building envelope. The aim of this project within IEA Task27 was to improve our knowledge on the complex situation and also to give a principal approach how to assess the performance of the building envelope. The participants have contributed to this aim not pretending that we have reached the end. (au)

  10. Panel discussion on near-field coupled processes with emphasis on performance assessment

    International Nuclear Information System (INIS)

    Codell, R.B.; Baca, R.G.; Ahola, M.P.

    1996-01-01

    The presentations in this panel discussion involve the general topic of near-field coupled processes and postclosure performance assessment with an emphasis on rock mechanics. The potential impact of near-field rock mass deformation on repository performance was discussed, as well as topics including long term excavation deterioration, the performance of geologic seals, and coupled processes concerning rock mechanics in performance assessments

  11. Assessment of backfill design for a KBS-3V repository: the BACLO program

    International Nuclear Information System (INIS)

    Jonsson, E.; Gunnarsson, D.; Hansen, J.; Keto, P.; Dixon, D.A.; Boergesson, L.

    2010-01-01

    Document available in extended abstract form only. Posiva and SKB initiated a joint programme BACLO (Backfilling and Closure of the Deep repository) in 2003 with the aim to develop methods and clay-based materials for backfilling the deposition tunnels of a repository utilizing the KBS-3V deposition concept. This paper summarises the work done in the third and final phase of the BACLO programme (2006-2008). The main objectives of this work were to examine backfill materials, deposition concepts and their importance to the clay-based block and pellet backfilling concept. Baclo Phase III was primarily intended to address the following four items: 1. evaluate options for design of block and pellet/granule materials for use in backfilling and in so doing provide a basis for selecting backfill materials; 2. provide a basis for recommending reference design(s) for backfilling through evaluation of materials, environmental processes and technical constraints likely to be encountered in a repository; 3. analyze how the potentially critical processes taking place during the installation and saturation phase affect the performance of the backfill and consequently the design basis for the backfill; and 4. evaluate how water will move through backfilled volumes and generally identify under what conditions water management will become an operational issue; and identify needs for further investigations and technical development. To address these objectives, studies were undertaken to examine how the various processes active during backfill installation and saturation as well as technical constraints affect its design basis. The work focused on the performance and technical feasibility of a block backfill concept, which calls for filling the majority of the tunnel volume with pre-compacted, clay-based backfill blocks and the remaining volume with bentonite pellets. Several backfill composition alternatives were chosen for study and they consisted of clay materials with differing

  12. Impacts of seismic activity on long-term repository performance at Yucca Mountain

    International Nuclear Information System (INIS)

    Gauthier, J.H.; Wilson, M.L.; Borns, D.J.; Arnold, B.W.

    1995-01-01

    Several effects of seismic activity on the release of radionuclides from a potential repository at Yucca Mountain are quantified. Future seismic events are predicted using data from the seismic hazard analysis conducted for the Exploratory Studies Facility (ESF). Phenomenological models are developed, including rockfall (thermal-mechanical and seismic) in unbackfilled emplacement drifts, container damage caused by fault displacement within the repository, and flow-path chance caused by changes in strain. Using the composite-porosity flow model (relatively large-scale, regular percolation), seismic events show little effect on total-system releases; using the weeps flow model (episodic pulses of flow in locally saturated fractures), container damage and flow-path changes cause over an order of magnitude increase in releases. In separate calculations using, more realistic representations of faulting, water-table rise caused by seismically induced changes in strain are seen to be higher than previously estimated by others, but not sufficient to reach a potential repository

  13. Performance Prediction for Large-Scale Nuclear Waste Repositories: Final Report

    International Nuclear Information System (INIS)

    Glassley, W E; Nitao, J J; Grant, W; Boulos, T N; Gokoffski, M O; Johnson, J W; Kercher, J R; Levatin, J A; Steefel, C I

    2001-01-01

    embodied in the code accurately represents the state-of-the-art in modeling these processes, and that the conceptualization of the models used in the simulations honors the primary processes that are controlling these systems. Application of the code to a wide range of important and strategic problems has been undertaken. Particularly significant are results obtained concerning the evolution of a potential high level nuclear waste repository at Yucca Mountain, Nevada. In these simulations, the results suggest that fluid movement and chemical changes will be such as to encourage solute transport around the sides of the waste emplacement tunnels, thus minimizing the potential for seepage of water and dissolved salts into the tunnels. The results also indicate that the short term response of the geological system to waste emplacement will be complex and rapid, and will be most readily detected below waste emplacement tunnels. A successful monitoring program of repository performance during the early stages of the operational period would thus benefit by coordinating design and execution of sampling strategies with a simulation tool such as NUFT-C. Such an approach would allow efficient and cost-effective sampling strategies, and would facilitate interpretation of what will surely be complex and massive data sets

  14. Methods for assessing the socioeconomic impacts of large-scale resource developments: implications for nuclear repository siting

    International Nuclear Information System (INIS)

    Murdock, S.H.; Leistritz, F.L.

    1983-03-01

    An overview of the major methods presently available for assessing the socioeconomic impacts of large-scale resource developments and includes discussion of the implications and applications of such methods for nuclear-waste-repository siting are provided. The report: (1) summarizes conceptual approaches underlying, and methodological alternatives for, the conduct of impact assessments in each substantive area, and then enumerates advantages and disadvantages of each alternative; (2) describes factors related to the impact-assessment process, impact events, and the characteristics of rural areas that affect the magnitude and distribution of impacts and the assessment of impacts in each area; (3) provides a detailed review of those methodologies actually used in impact assessment for each area, describes advantages and problems encountered in the use of each method, and identifies the frequency of use and the general level of acceptance of each technique; and (4) summarizes the implications of each area of projection for the repository-siting process, the applicability of the methods for each area to the special and standard features of repositories, and makes general recommendations concerning specific methods and procedures that should be incorporated in assessments for siting areas

  15. Preparing for a Trustworthiness Assessment of the National Transportation Library’s Digital Repository ROSA P

    Science.gov (United States)

    2018-01-01

    The National Transportation Library (NTL) is an all-digital repository of transportation knowledge that falls under federal mandates to serve as a central clearinghouse for transportation data and information of the Federal Government. as well ...

  16. 10 CFR 63.113 - Performance objectives for the geologic repository after permanent closure.

    Science.gov (United States)

    2010-01-01

    ... closure. (a) The geologic repository must include multiple barriers, consisting of both natural barriers... in combination with natural barriers, radiological exposures to the reasonably maximally exposed... engineered barrier system must be designed so that, working in combination with natural barriers, releases of...

  17. Role of groundwater oxidation potential and radiolysis on waste glass performance in crystalline repository environments

    International Nuclear Information System (INIS)

    Jantzen, C.M.; Bibler, N.E.

    1985-01-01

    Laboratory experiments have shown that groundwater conditions in a Stripa granite repository will be as reducing as those in a basalt repository. The final oxidation potential (Eh) at 70 0 C for Stripa groundwater deaerated and equilibrated with crystalline granite was -0.45V. In contrast, the oxidation potential at 60 0 C for Grande Ronde groundwater equilibrated with basalt was -0.40V. The reducing groundwater conditions were found to slightly decrease the time-dependent release of soluble components from the waste glass. Spectrophotometric analysis of the equilibrated groundwaters indicated the presence of Fe 2+ confirming that the Fe 2+ /Fe 3+ couple is controlling the oxidation potential. It was also shown that in the alkaline pH regime of these groundwaters the iron species are primarily associated with x-ray amorphous precipitates in the groundwater. Gamma radiolysis in the absence of waste glass and in the absence of oxygen further reduces the oxidation potential of both granitic and basaltic groundwaters. The effect is more pronounced in the basaltic groundwater. The mechanism for this decrease is under investigation but appears related to the reactive amorphous precipitate. The results of these tests suggest that H 2 may not escape from the repository system as postulated and that radiolysis may not cause the groundwaters to become oxidizing in a crystalline repository when abundant Fe 2+ species are present. 23 refs., 3 figs., 3 tabs

  18. Long Term Behaviour of 14C and Stability Assessments of Graphite Under Repository Conditions

    International Nuclear Information System (INIS)

    Jones, Abbie N.; McDermott, Lorraine; Worth, Robert; Hagos, Bereket; Black, Greg; Marsden, Barry J

    2016-01-01

    The key objectives of the University of Manchester’s nuclear graphite research within the CRP are to provide analysis on the long term behaviour and stability assessments of irradiated graphite waste. The research will concentrate on isotopic 14 C mobility under repository environments. This also requires an understanding the long-term behaviour of the final waste form under repository conditions. Procedures to evaluate the long term leaching properties of radionuclides from irradiated graphite waste has been developed by combining ANSI 16.1 (USA) and NEN 7345 (Netherlands) standardised diffusion leaching techniques. The ANSI 16.1 standard has been followed to acquire the leachates and to determine the leach rate and diffusion coefficient. The NEN 7345 standard technique has been used to determine the diffusion mechanism of radionuclides. The investigation employs simulated Drigg groundwater as a leachant using semi-dynamic technique for the production of leachate specimens. Analysis of 3 H and 14 C activity release from Magnox graphite was measured using liquid scintillating counting. Preliminary results show that there is an initial high release of activity and decreases when the leaching period increases. This may be due to the depletion of contaminants that were initially bound by the internal pore networks and the free surface. During the leaching test approximately 275.33 ± 18.20 Bq of 3 H and 106.26 ± 7.01 Bq of 14 C was released into the leachant within 91 days. The work reported herein contributed several key findings to the international work on graphite leaching to offer guidance leading toward obtaining leaching data in the future: (a) the effective diffusion coefficient for 14 C from graphite waste has been determined. The diffusion process for 14 C has two stages resulting two different values of diffusion coefficient, i.e., for the fast and slow components; (b) the controlling leaching mechanism for 3 H radionuclide from graphite is shown to be

  19. Regional Groundwater Flow Assessment in a Prospective High-Level Radioactive Waste Repository of China

    Directory of Open Access Journals (Sweden)

    Xiaoyuan Cao

    2017-07-01

    Full Text Available The production of nuclear energy will result in high-level radioactive waste (HLRW, which brings potential environmental dangers. Selecting a proper disposal repository is a crucial step in the development of nuclear energy. This paper introduces firstly the hydrogeological conditions of the Beishan area in China. Next, a regional groundwater model is constructed using a multiphase flow simulator to analyze the groundwater flow pattern in the Beishan area. Model calibration shows that the simulated and observed hydraulic heads match well, and the simulated regional groundwater flow pattern is similar to the surface flow pattern from the channel network, indicating that the groundwater flow is mainly dependent on the topography. In addition, the simulated groundwater storage over the period from 2003 to 2014 is similar to the trend derived from the Gravity Recovery and Climate Experiment satellite-derived results. Last, the established model is used to evaluate the influences of the extreme climate and regional faults on the groundwater flow pattern. It shows that they do not have a significant influence on the regional groundwater flow patterns. This study will provide a preliminary reference for the regional groundwater flow assessment in the site of the HLRW in China.

  20. Biosphere Modeling for the Dose Assessment of a HLW Repository: Development of ACBIO

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-01-15

    For the purpose of evaluating a dose rate to an individual due to a long-term release of nuclides from a HLW repository, a biosphere assessment model and an implemented code, ACBIO, based on the BIOMASS methodology have been developed by utilizing AMBER, a general compartment modeling tool. To demonstrate its practicability and usability as well as to observe the sensitivity of the compartment scheme, the concentration, the activity in the compartments as well as the annual flux between the compartments at their peak values, were calculated and investigated. For each case when changing the structure of the compartments and GBIs as well as varying selected input Kd values, all of which seem very important among the others, the dose rate per nuclide release rate is calculated separately and analyzed. From the maximum dose rates, the flux to dose conversion factors for each nuclide were derived, which are used for converting the nuclide release rate appearing from the geosphere through various GBIs to dose rates (Sv/y) for an individual in a critical group. It has also been observed that the compartment scheme, the identification of a possible exposure group and the GBIs could all be highly sensitive to the final consequences in a biosphere modeling.

  1. Biosphere Modeling for the Dose Assessment of a HLW Repository: Development of ACBIO

    International Nuclear Information System (INIS)

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

    2006-01-01

    For the purpose of evaluating a dose rate to an individual due to a long-term release of nuclides from a HLW repository, a biosphere assessment model and an implemented code, ACBIO, based on the BIOMASS methodology have been developed by utilizing AMBER, a general compartment modeling tool. To demonstrate its practicability and usability as well as to observe the sensitivity of the compartment scheme, the concentration, the activity in the compartments as well as the annual flux between the compartments at their peak values, were calculated and investigated. For each case when changing the structure of the compartments and GBIs as well as varying selected input Kd values, all of which seem very important among the others, the dose rate per nuclide release rate is calculated separately and analyzed. From the maximum dose rates, the flux to dose conversion factors for each nuclide were derived, which are used for converting the nuclide release rate appearing from the geosphere through various GBIs to dose rates (Sv/y) for an individual in a critical group. It has also been observed that the compartment scheme, the identification of a possible exposure group and the GBIs could all be highly sensitive to the final consequences in a biosphere modeling

  2. Use of Geohydrologic Response Functions in the assessment of deep nuclear waste repositories: assessment of effectiveness of geologic isolation systems

    International Nuclear Information System (INIS)

    Nelson, R.W.

    1981-05-01

    Geohydrologic Response Functions (GRFs) interrelate the three vital factors needed in the repository decision-making process: the quantity, arrival time, and location of contamination reaching the biosphere. GRFs further focus attention upon two related and additive parameters: the initial delay time and delay spread time. After the principal site selection, the GRFs may be applied to obtain more detailed performance evaluations concerning specific nuclides or waste components, specific results for various types of accidental release, and effects of a variety of contaminant source terms or leach models. The response functions may be applied to consider contaminant reductions and delays through material sorption as well as through a variety of other interactions and effects

  3. Deep geologic disposal. Lessons learnt from recent performance assessment studies

    International Nuclear Information System (INIS)

    Pescatore, C.; Andersson, J.

    1998-01-01

    Performance assessment (PA) studies are part of the decision basis for the siting, operation, and closure of deep repositories of long-lived nuclear wastes. In 1995 the NEA set up the Working Group on Integrated Performance Assessments of Deep Repositories (IPAG) with the goals to analyse existing PA studies, learn about what has been produced to date, and shed light on what could be done in future studies. Ten organisations submitted their most recent PA study for analysis and discussion, including written answers to over 70 questions. Waste management programmes, disposal concepts, geologies, and different types and amounts of waste offered a unique opportunity for exchanging information, assessing progress in PA since 1990, and identifying recent trends. A report was completed whose main lessons are overviewed. (author)

  4. 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)

  5. Radioactive waste repository study

    International Nuclear Information System (INIS)

    1978-11-01

    This is the second part of a report of a preliminary study for AECL. It considers the requirements for an underground waste repository for the disposal of wastes produced by the Canadian Nuclear Fuel Program. The following topics are discussed with reference to the repository: 1) geotechnical assessment, 2) hydrogeology and waste containment, 3) thermal loading and 4) rock mechanics. (author)

  6. Potential impact of Andrassy bentonite microbial diversity in the long-term performance of a deep nuclear waste repository

    Science.gov (United States)

    Tadza, M. Y. Mohd; Tadza, M. A. Mohd; Bag, R.; Harith, N. S. H.

    2018-01-01

    Copper and steel canning and bentonite buffer are normally forseen as the primary containment component of a deep nuclear waste repository. Distribution of microbes in subsurface environments have been found to be extensive and directly or indirectly may exert influence on waste canister corrosion and the mobility of radionuclides. The understanding of clays and microbial interaction with radionuclides will be useful in predicting the microbial impacts on the performance of the waste repositories. The present work characterizes the culture-dependent microbial diversity of Andrassy bentonite recovered from Tawau clay deposits. The evaluation of microbial populations shows the presence of a number of cultivable microbes (e.g. Staphylococcus, Micrococcus, Achromobacter, Bacillus, Paecilomyces, Trichoderma, and Fusarium). Additionally, a pigmented yeast strain Rhodotorula mucilaginosa was also recovered from the formation. Both Bacillus and Rhodotorula mucilaginosa have high tolerance towards U radiation and toxicity. The presence of Rhodotorula mucilaginosa in Andrassy bentonite might be able to change the speciation of radionuclides (e.g. uranium) in a future deep repository. However, concern over the presence of Fe (III) reduction microbes such as Bacillus also found in the formation could lead to corrosion of copper steel canister and affect the overall performance of the containment system.

  7. 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

  8. Comment on the internal consistency of thermodynamic databases supporting repository safety assessments

    International Nuclear Information System (INIS)

    Arthur, R.C.

    2001-11-01

    This report addresses the concept of internal consistency and its relevance to the reliability of thermodynamic databases used in repository safety assessments. In addition to being internally consistent, a reliable database should be accurate over a range of relevant temperatures and pressures, complete in the sense that all important aqueous species, gases and solid phases are represented, and traceable to original experimental results. No single definition of internal consistency need to be universally accepted as the most appropriate under all conditions, however. As a result, two databases that are each internally consistent may be inconsistent with respect to each other, and a database derived from two or more such databases must itself be internally inconsistent. The consequences of alternative definitions that are reasonably attributable to the concept of internal consistency can be illustrated with reference to the thermodynamic database supporting SKB's recent SR 97 safety assessment. This database is internally inconsistent because it includes equilibrium constants calculated over a range of temperatures: using conflicting reference values for some solids, gases and aqueous species that are common to two internally consistent databases (the OECD/NEA database for radioelements and SUPCRT databases for non-radioactive elements) that serve as source databases for the SR 97 TDB, using different definitions in these source databases of standard states for condensed phases and aqueous species, based on different mathematical expressions used in these source databases representing the temperature dependence of the heat capacity, and based on different chemical models adopted in these source databases for the aqueous phase. The importance of such inconsistencies must be considered in relation to the other database reliability criteria noted above, however. Thus, accepting a certain level of internal inconsistency in a database it is probably preferable to use a

  9. Comment on the internal consistency of thermodynamic databases supporting repository safety assessments

    Energy Technology Data Exchange (ETDEWEB)

    Arthur, R.C. [Monitor Scientific, LLC, Denver, CO (United States)

    2001-11-01

    This report addresses the concept of internal consistency and its relevance to the reliability of thermodynamic databases used in repository safety assessments. In addition to being internally consistent, a reliable database should be accurate over a range of relevant temperatures and pressures, complete in the sense that all important aqueous species, gases and solid phases are represented, and traceable to original experimental results. No single definition of internal consistency need to be universally accepted as the most appropriate under all conditions, however. As a result, two databases that are each internally consistent may be inconsistent with respect to each other, and a database derived from two or more such databases must itself be internally inconsistent. The consequences of alternative definitions that are reasonably attributable to the concept of internal consistency can be illustrated with reference to the thermodynamic database supporting SKB's recent SR 97 safety assessment. This database is internally inconsistent because it includes equilibrium constants calculated over a range of temperatures: using conflicting reference values for some solids, gases and aqueous species that are common to two internally consistent databases (the OECD/NEA database for radioelements and SUPCRT databases for non-radioactive elements) that serve as source databases for the SR 97 TDB, using different definitions in these source databases of standard states for condensed phases and aqueous species, based on different mathematical expressions used in these source databases representing the temperature dependence of the heat capacity, and based on different chemical models adopted in these source databases for the aqueous phase. The importance of such inconsistencies must be considered in relation to the other database reliability criteria noted above, however. Thus, accepting a certain level of internal inconsistency in a database it is probably preferable to

  10. The Swedish final repository for reactor waste (SFR). A summary of the SFR project with special emphasis on the near-field assessments

    International Nuclear Information System (INIS)

    Carlsson, J.

    1988-01-01

    The first phase of the final repository for reactor waste (SFR) is scheduled for operation in April 1988. The construction work is finished and preoperational tests are in progress. Impact on the environment from SFR is analysed in a final safety report. This paper gives a summary of the design and performance of SFR. Assessments, made for the analysises of the long term safety, are given with special emphasis on the near-field. As a conclusion from the analysises, the dose commitment to the most affected individual during the post-closure period, has proved to constitute only an insignificant contribution to the natural radioactive environment of the area

  11. Zedex - an in-situ study of the importance of the excavation disturbed zone to repository performance

    International Nuclear Information System (INIS)

    Olsson, O.; Slimane K.B.; Davies, N.

    1995-01-01

    The excavation of shafts and tunnels in a waste repository will cause a disturbance to the surrounding rock mass with possible alterations to rock mass stability and hydraulic properties. The properties and extent of the disturbed zone must be considered in the design of a repository and in the assessment of its long-term safety. In addition, the data collected in drifts, which will be used for detailed characterization of the repository, will be affected by excavation effects. ANDRA, UK Nirex, and SKB have jointly set up the ZEDEX Project at the Aespoe Hard Rock Laboratory, Sweden, in order to improve understanding of the properties and extent the Excavation Disturbed Zone (EDZ). The main objective is to study the mechanical behavior of the EDZ with respect to its origin, character, magnitude of property change, extent, and its dependence of excavation method. The project will also provide a test of equipment and methodology for quantifying the EDZ. The ZEDEX Project includes tests of open-quote normal close-quote excavation by drill and blast, a smooth blasting procedure based on low shock explosives, and excavation by tunnel boring. The ZEDEX Project is still in progress and results have not been fully analyzed. The preliminary analysis of the results indicates that the measurable changes in properties induced by tunnel excavation are comparatively small

  12. The impact of repository heat on thermo-hydrological performance at Yucca Mountain

    International Nuclear Information System (INIS)

    Buscheck, T.A.; Nitao, J.J.

    1993-09-01

    To safely and permanently store high-level nuclear waste, the potential Yucca Mountain repository site must mitigate the release and transport of radionuclides for tens of thousands of years. In the failure scenario of greatest concern, water would contact a waste package (WP), accelerate its failure rate, and eventually transport radionuclides to the water table. These analyses have demonstrated that the only significant source of liquid water is nonequilibrium fracture flow from: (1) meteoric sources, (2) condensate drainage generated under boiling conditions, and (3) condensate drainage generated under sub-boiling conditions. The first source of liquid water arises from the ambient system; the second and third sources are generated by repository heat. Buoyant vapor flow, occurring either on a sub-repository scale or on a mountain scale, may play an important role in the generation of the second and third sources of liquid water. By considering a wide range in bulk permeability, k b , the authors identify the threshold k b (called k b hyd ) at which buoyant, vapor convection begins to dominate hydrological behavior, and the threshold k b (called k b th ) at which this convection begins to dominate thermal behavior. They find that k b th is generally an order of magnitude larger than k b hyd and that the development of a large above-boiling zone suppresses the effects of buoyant vapor flow. Of particular concern are conditions that promote the focusing of vapor flow and condensate drainage, which could result in persistent two-phase conditions (often referred to as the heat-pipe effect) in the vicinity of WPs. The results of this study underscore the need for in situ heater tests to help diagnose the potential for the major repository-heat-driven sources of fracture flow

  13. An assessment of the radiological impact of coastal erosion of the UK Low-Level Waste Repository - 59137

    International Nuclear Information System (INIS)

    Sumerling, Trevor; Shevelan, John; Cummings, Richard; Fish, Paul; Towler, George; Penfold, James

    2012-01-01

    The UK Low Level Waste Repository Ltd submitted an Environmental Safety Case for the disposal of low-level waste to our regulator, the Environment Agency, on the 1 May 2011. This includes assessments of the long-term radiological safety of past and future disposals. A particular feature of the Low Level Waste Repository (LLWR) is that, because of its proximity to the coast, the site is vulnerable to coastal erosion. Our present understanding is that the site will be eroded on a timescale of a few hundred to a few thousand years, with consequent disruption of the repository, and dispersal of the wastes. We have undertaken a programme of scientific research and monitoring to characterise the evolution and function of the current coastal system that provides a basis for forecasting its future evolution. This has included modelling of contemporary hydrodynamics, geo-morphological mapping, repeat LiDAR and aerial photographic surveys to detect patterns and rates of change, coastal inspections and reconstructions of post-glacial (i.e. last 15, 000 years) sea levels and sediment budgets. Estimates of future sea-level rise have been derived from international sources and consideration given to the impact of such on the local coastline. Two alternative models of coastal recession have then been applied, one empirical and one physical-process based, taking account of the composition of Quaternary-age sediments between the coast and the site and uncertainties in future local sea level change. Comparison of the ranges of calculated times to site contact with sea-level rise indicate that the repository is most likely to be disrupted by undercutting of the engineered vaults and of the trenches. A novel and flexible radiological assessment model has been developed to analyse the impacts of the erosion of the repository and subsequent dispersal of wastes. The model represents the spatial layout of the site and distribution of radionuclides within the repository and is able to take

  14. Colloid-Facilitated Radionuclide Transport: Current State of Knowledge from a Nuclear Waste Repository Risk Assessment Perspective

    International Nuclear Information System (INIS)

    Reimus, Paul William; Zavarin, Mavrik; Wang, Yifeng

    2017-01-01

    This report provides an overview of the current state of knowledge of colloid-facilitated radionuclide transport from a nuclear waste repository risk assessment perspective. It draws on work that has been conducted over the past 3 decades, although there is considerable emphasis given to work that has been performed over the past 3-5 years as part of the DOE Used Fuel Disposition Campaign. The timing of this report coincides with the completion of a 3-year DOE membership in the Colloids Formation and Migration (CFM) partnership, an international collaboration of scientists studying colloid-facilitated transport of radionuclides at both the laboratory and field-scales in a fractured crystalline granodiorite at the Grimsel Test Site in Switzerland. This Underground Research Laboratory has hosted the most extensive and carefully-controlled set of colloid-facilitated solute transport experiments that have ever been conducted in an in-situ setting, and a summary of the results to date from these efforts, as they relate to transport over long time and distance scales, is provided in Chapter 3 of this report.

  15. Colloid-Facilitated Radionuclide Transport: Current State of Knowledge from a Nuclear Waste Repository Risk Assessment Perspective

    Energy Technology Data Exchange (ETDEWEB)

    Reimus, Paul William [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Zavarin, Mavrik [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Wang, Yifeng [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2017-01-25

    This report provides an overview of the current state of knowledge of colloid-facilitated radionuclide transport from a nuclear waste repository risk assessment perspective. It draws on work that has been conducted over the past 3 decades, although there is considerable emphasis given to work that has been performed over the past 3-5 years as part of the DOE Used Fuel Disposition Campaign. The timing of this report coincides with the completion of a 3-year DOE membership in the Colloids Formation and Migration (CFM) partnership, an international collaboration of scientists studying colloid-facilitated transport of radionuclides at both the laboratory and field-scales in a fractured crystalline granodiorite at the Grimsel Test Site in Switzerland. This Underground Research Laboratory has hosted the most extensive and carefully-controlled set of colloid-facilitated solute transport experiments that have ever been conducted in an in-situ setting, and a summary of the results to date from these efforts, as they relate to transport over long time and distance scales, is provided in Chapter 3 of this report.

  16. Role of groundwater oxidation potential and radiolysis on waste glass performance in crystalline repository environments

    International Nuclear Information System (INIS)

    Jantzen, C.M.; Bibler, N.E.

    1986-01-01

    Laboratory experiments have shown that groundwater conditions in a granite repository will be as reducing as those in a basalt repository. Chemical analysis of the reduced groundwaters confirmed that the Fe 2+ /Fe 3+ couple controls the oxidation potential (Eh). The reducing groundwater conditions were found to decrease the time-dependent release of soluble elements (Li and B) from the waste glass. However, due to the lower solubility of multivalent elements released from the glass when the groundwaters are reducing, these elements have significantly lower concentrations in the leachates. Gamma radiolysis reduced the oxidation potential of both granitic and basaltic groundwater in the absence of both waste glass and oxygen. This occurred in tests at atmospheric pressure where H 2 could have escaped from the solution. The mechanism for this decrease in Eh is under investigation but appears related to the reactive amorphous precipitate in both groundwaters. The results of these tests suggest that radiolysis may not cause the groundwaters to become oxidizing in a crystalline repository when abundant Fe 2+ species are present

  17. Assessment of effectiveness of geologic isolation systems. Test case release consequence analysis for a spent fuel repository in bedded salt

    International Nuclear Information System (INIS)

    Raymond, J.R.; Bond, F.W.; Cole, C.R.; Nelson, R.W.; Reisenauer, A.E.; Washburn, J.F.; Norman, N.A.; Mote, P.A.; Segol, G.

    1980-01-01

    Geologic and geohydrologic data for the Paradox Basin have been used to simulate movement of ground water and radioacrtive contaminants from a hypothetical nuclear reactor spent fuel repository after an assumed accidental release. The pathlines, travel times and velocity of the ground water from the repository to the discharge locale (river) were determined after the disruptive event by use of a two-dimensional finite difference hydrologic model. The concentration of radioactive contaminants in the ground water was calculated along a series of flow tubes by use of a one-dimensional mass transport model which takes into account convection, dispersion, contaminant/media interactions and radioactive decay. For the hypothetical site location and specific parameters used in this demonstration, it is found that Iodine-129 (I-129) is tthe only isotope reaching the Colorado River in significant concentration. This concentration occurs about 8.0 x 10 5 years after the repository has been breached. This I-129 ground-water concentration is about 0.3 of the drinking water standard for uncontrolled use. The groundwater concentration would then be diluted by the Colorado River. None of the actinide elements reach more than half the distance from the repository to the Colorado River in the two-million year model run time. This exercise demonstrates that the WISAP model system is applicable for analysis of contaminant transport. The results presented in this report, however, are valid only for one particular set of parameters. A complete sensitivity analysis must be performed to evaluate the range of effects from the release of contaminants from a breached repository

  18. Safety assessment for a potential SNF repository and its implication to the proliferation resistance nuclear fuel cycle

    International Nuclear Information System (INIS)

    Hwang, Y.; Jeong, M.S.; Seo, C.S.

    2007-01-01

    KAERI is developing the pyro-process technology to minimize the burden on permanent disposal of spent nuclear fuel. In addition, KAERI has developed the Korean Reference System for potential spent nuclear fuel disposal since 1997. The deep geologic disposal system is composed of a multi-barrier system in a crystalline rock to dispose of 36,000 MT of spent nuclear fuel (SNF) from a CANDU and a PWR. Quite recently, introduction of advanced nuclear fuel cycles such as pyro-processing is a big issue to solve the everlasting disposal problem and to assure the sustainable supply of fuel for reactors. To compare the effect of direct disposal of SNF with that of the high level waste disposal for waste generated from the advanced nuclear fuel cycles, the total system performance assessment for two different schemes is developed; one for direct disposal of SNF and the other for the introduction of the pyro-processing and direct disposal CANDU spent nuclear fuel. The safety indicators to assess the environmental friendliness of the disposal option are annual individual doses, toxicities and risks. Even though many scientists use the toxicity to understand the environmental friendliness of the disposal, scientifically the annual individual doses or risks are meaningful indicators for it. The major mechanisms to determine the doses and risks for direct disposal are as follows: (1) Dissolution mechanisms of uranium dioxides which control the dissolution of most nuclides such as TRU's and most parts of fission products. (2) Instant release fraction of highly soluble nuclides such as I-129, C-135, Tc-99, and others. (3) Retardation and dilution effect of natural and engineered barriers. (4) Dilution effect in the biosphere. The dominant nuclide is I-129 which follows both congruent and instantaneous release modes. Since its long half life associated with the instantaneous release I-129 is dominant well beyond one million. The impact of the TRU's is negligible until the significant

  19. 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

  20. Staged Repository Development Programmes

    International Nuclear Information System (INIS)

    Isaacs, T

    2003-01-01

    Programs to manage and ultimately dispose of high-level radioactive wastes are unique from scientific and technological as well as socio-political aspects. From a scientific and technological perspective, high-level radioactive wastes remain potentially hazardous for geological time periods-many millennia-and scientific and technological programs must be put in place that result in a system that provides high confidence that the wastes will be isolated from the accessible environment for these many thousands of years. Of course, ''proof'' in the classical sense is not possible at the outset, since the performance of the system can only be known with assurance, if ever, after the waste has been emplaced for those geological time periods. Adding to this challenge, many uncertainties exist in both the natural and engineered systems that are intended to isolate the wastes, and some of the uncertainties will remain regardless of the time and expense in attempting to characterize the system and assess its performance. What was perhaps underappreciated in the early days of waste management and repository program development were the unique and intense reactions that the institutional, political, and public bodies would have to repository program development, particularly in programs attempting to identify and then select sites for characterization, design, licensing, and ultimate development. Reactions in most nations were strong, focused, unrelenting, and often successful in hindering, derailing, and even stopping national repository programs. The reasons for such reactions and the measures to successfully respond to them are still evolving and continue to be the focus of many national program and political leaders. Adaptive Staging suggests an approach to repository program development that reflects the unique challenges associated with the disposal of high-level radioactive waste. The step-wise, incremental, learn-as-you-go approach is intended to maximize the

  1. 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

  2. Probabilistic modelling of gas generation in nuclear waste repositories under consideration of new studies performed at the WIPP

    International Nuclear Information System (INIS)

    Niemeyer, M.; Wilhelm, S.; Poppei, J.

    2012-01-01

    Document available in extended abstract form only. The inventory of a nuclear waste repository includes significant amounts of metal and organic matter. Under the prevailing conditions in a repository in a salt formation in contact with water, these materials tend to react and transform under significant gas production. This increases the pressure and potentially leads to an enhanced transport of radio nuclides. Therefore, these phenomena need to be understood and characterized in detail for the assessment of the safety of the repository A modelling code, GASGEN, developed by AF-Consult Switzerland Ltd to predict the evolution of gas production by microbial processes and anaerobic corrosion of metal, was applied at two locations of repositories in salt rock in Germany. Therein, the microbial decomposition of organic waste components is modelled by the sub-processes of denitrification, reduction of sulphates, fermentation and methano-genesis. The models differentiate between highly degradable cellulose and materials of lesser degradability, such as polymers. Gas production through anaerobic corrosion of metal is mainly due to the iron content of the waste. In addition, the precipitation of carbonate from alkaline materials in the inventory (e.g. cement) is considered. The inventories of contained waste, which determine the amount of gas that can be produced, are subject to uncertainties. The rates of the various reactions also depend on numerous factors and are therefore variable. In order to cover this variability, gas production is modelled probabilistically. In this way the behaviour of the gas generation can be estimated together with its bandwidth Figure 1. In addition to the produced amounts of gas, the model calculations also consider the potential of acidification of the fluid enclosed in the repository chambers and the effect of a changing pH-level on the rate of corrosion. Based on results, the effect of a pH-dependent corrosion rate is illustrated and the

  3. Survey and review of near-field performance assessment

    International Nuclear Information System (INIS)

    Apted, M.J.

    1993-01-01

    Chemical reactions control the performance, stability, and rate of degradation of natural and engineered barriers to waste repositories of the near field. Chemical processes are overviewed in this context. Temperature, and associated temperature gradients, are also important parameters in near-field performance assessment. The mechanical conditions of the near-field rock will be perturbed by construction of the underground repository. Mechanical analysis in the near field is further complicated by the introduction of HLW canisters and associated engineered barrier materials. Hydrological processes important to near-field performance include those associated with fluid transport. Considerable discussions and studies have been conducted on the issue of coupling among chemical-thermal-mechanical-hydrological processes; they are overviewed. (R.P.) 2 figs., 2 tabs

  4. Evaluation of repository safety

    Energy Technology Data Exchange (ETDEWEB)

    Sagar, B.; Patrick, W.; Dasgupta, B.; Mohanty, S. [Center for Nuclear Waste Regulatory Analyses, San Antonio (United States)

    2002-07-01

    . Mathematical models play an even larger role in evaluating repository safety in the much longer postclosure period. During this period, repository performance is evaluated considering gradual degradation of engineered barriers, together with possible slow changes in the natural system (e.g., climate) and under conditions of potential discrete and sudden disruptive events (e.g., volcanic eruption, seismic ground motion, and direct fault movement). The general aim of postclosure performance assessment models is to simulate the future behavior of the repository in a manner that is sufficiently simplified to be tractable, yet sufficiently realistic to give reasonable (or bounded) estimates of risk to future generations. The simplifications are based on information gained by using process level models, natural analog studies, and laboratory and fieldwork. Because of large uncertainties inherent in characterizing a large and complex system for such long periods, probabilistic simulations are generally preferred. In this paper, we briefly describe the preclosure and postclosure safety evaluation models developed jointly by the Center for Nuclear Waste Regulatory Analyses and the U.S. Nuclear Regulatory Commission. We give examples illustrating how we intend to use these models within the regulatory framework to evaluate the required.

  5. Evaluation of repository safety

    International Nuclear Information System (INIS)

    Sagar, B.; Patrick, W.; Dasgupta, B.; Mohanty, S.

    2002-01-01

    . Mathematical models play an even larger role in evaluating repository safety in the much longer postclosure period. During this period, repository performance is evaluated considering gradual degradation of engineered barriers, together with possible slow changes in the natural system (e.g., climate) and under conditions of potential discrete and sudden disruptive events (e.g., volcanic eruption, seismic ground motion, and direct fault movement). The general aim of postclosure performance assessment models is to simulate the future behavior of the repository in a manner that is sufficiently simplified to be tractable, yet sufficiently realistic to give reasonable (or bounded) estimates of risk to future generations. The simplifications are based on information gained by using process level models, natural analog studies, and laboratory and fieldwork. Because of large uncertainties inherent in characterizing a large and complex system for such long periods, probabilistic simulations are generally preferred. In this paper, we briefly describe the preclosure and postclosure safety evaluation models developed jointly by the Center for Nuclear Waste Regulatory Analyses and the U.S. Nuclear Regulatory Commission. We give examples illustrating how we intend to use these models within the regulatory framework to evaluate the required

  6. Georgia's Teacher Performance Assessment

    Science.gov (United States)

    Fenton, Anne Marie; Wetherington, Pamela

    2016-01-01

    Like most states, Georgia until recently depended on an assessment of content knowledge to award teaching licenses, along with a licensure recommendation from candidates' educator preparation programs. While the content assessment reflected candidates' grasp of subject matter, licensure decisions did not hinge on direct, statewide assessment of…

  7. The enhancement of a biosphere code for use in the assessment of deep repositories for radioactive waste

    International Nuclear Information System (INIS)

    Ashton, J.; Little, R.H.

    1991-01-01

    A disposal system for radioactive waste is conventionally considered to consist of the engineered barriers of the repository, the geosphere and the surface environment or biosphere. Computer codes have been developed to assist in assessing the impact of radionuclides migrating from the repository through the disposal system. Codes have been developed to represent the repository, the geosphere and the biosphere. The fundamental role of the biosphere codes is, for radionuclide inputs, to estimate the dose, or probability distribution function of dose, to a maximally exposed individual as a function of time. In the United Kingdom, the primary target for long-term radiological impacts from a single disposal facility is that the risk of fatal cancer to an individual in any one year should not exceed one in a million at any time. The recent revival of interest in the UK in the deep disposal of radioactive waste has resulted in the need for the risk to be estimated over timescales up to a million years, since a deep geologic repository might be able to withstand the effects of several future glacial episodes. Biosphere modelling for deep disposal of radioactive waste poses particular problems since the surface environment is expected to evolve as a result of changes in climatic conditions. Consequently the effect of climate induced changes in geomorphology, land use and sea level on the calculated risk have to be considered. This paper outlines the development of a new version of the dynamic biosphere model DECOS, which was developed originally in the context of shallow site assessments. The new version of the code, called DECOS-MG, is capable of simulating the effect of multiple glacial cycles and changes in sea level. (14 refs., 3 figs., 2 tabs.)

  8. The Case for Standardized Reporting and Assessment Requirements for Institutional Repositories

    Science.gov (United States)

    McDonald, Robert H.; Thomas, Chuck

    2008-01-01

    Research libraries traditionally have focused their efforts on building local collections to support the research needs of their organizations and providing for the discovery and use of resources within these collections. The emergence in recent years of disciplinary and institutional digital repositories presents new challenges for research…

  9. Repositories of Open Educational Resources: An Assessment of Reuse and Educational Aspects

    Science.gov (United States)

    Santos-Hermosa, Gema; Ferran-Ferrer, Núria; Abadal, Ernest

    2017-01-01

    This article provides an overview of the current state of repositories of open educational resources (ROER) in higher education at international level. It analyses a series of educational indicators to determine whether ROER can meet the specific needs of the education context, and to clarify understanding of the reuse of open educational…

  10. 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

  11. 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

  12. Biosphere modelling for the assessment of radioactive waste repositories; the development of a common basis by the BIOMOVS II reference biospheres working group

    International Nuclear Information System (INIS)

    Dorp, F. van; Egan, M.; Kessler, J.H.; Nilsson, S.; Pinedo, P.; Smith, G.; Torres, C.

    1998-01-01

    Performance criteria for radioactive waste repositories are often expressed in terms of dose or risk. The characteristics of biosphere modelling for performance assessment are that: (a) potential release occurs in the distant future, (b) reliable predictions of human behaviour at the time of release are impracticable, and (c) the biosphere is not considered to be a barrier as the geosphere and the engineered barriers. For these and other reasons, differences have arisen in the approaches to biosphere modelling for repository dose and risk assessment. The BIOMOVS II Reference Biospheres Working Group has developed (a) a recommended methodology for biosphere model development, (b) a structured list of features, events and processes (FEPs) which the model should describe, and (c) an illustrative example of the recommended methodology. The Working Group has successfully tested the Interaction Matrix (or Rock Engineering Systems, RES) approach for developing conceptual models. The BIOMOVS II Working Groups on Reference Biospheres and Complementary Studies have laid the basis for considerable harmonisation in approaches to biosphere modelling of long term radionuclide releases. (Copyright (c) 1998 Elsevier Science B.V., Amsterdam. All rights reserved.)

  13. Ontario Power Generation's proposed L and ILW deep geologic repository: geo-scientific assessment

    International Nuclear Information System (INIS)

    Jensen, Mark; Raven, Ken; Leech, Robert

    2012-01-01

    Document available in extended abstract form only. The Nuclear Waste Management Organization (NWMO) on behalf of Ontario Power Generation (OPG) has conducted multi-disciplinary geo-scientific studies at the Bruce nuclear site to confirm the suitability of the site to host a proposed Deep Geologic Repository (DGR) for the long-term management of Low and Intermediate Level Radioactive Waste (L and ILW) from OPG owned or operated nuclear generating facilities. An Environmental Assessment for the proposed DGR is currently underway in accordance with the Canadian Environmental Assessment Act. Bruce nuclear site, situated 225 km northwest of Toronto on the eastern shore of Lake Huron, is underlain by an 850 m thick sedimentary sequence of Cambrian to Devonian age near-horizontally layered, weakly deformed shales, carbonates and evaporites of the Michigan Basin. Within this sedimentary pile, the proposed DGR would be excavated within the low permeability argillaceous limestone of the Cobourg Formation at a depth of 680 m, which is overlain by 200 m of Upper Ordovician shale formations (Figure 1). A key aspect of the DGR Safety Case is the predictable nature and long-term stability and integrity of the sedimentary sequence to contain and isolate L and ILW at time frames on the order of 1 Ma. Early in the project, geo-scientific studies that considered regional and site-specific public domain data sets indicated favourable geologic conditions for implementation of the DGR concept (Golder, 2003; Mazurek, 2004). Geo-scientific studies for the DGR were initiated in 2006 following decision by the Municipality of Kincardine to support the DGR concept. Geo-scientific activities were divided into two key areas; i) Site-specific geo-scientific studies of the Bruce nuclear site (i.e., Descriptive Geosphere Site Model); and ii) a Geo-synthesis to convey an understanding of past, present and future evolution of Geosphere enclosing the DGR relevant to communicating notions of

  14. Postclosure performance assessment of the SCP [Site Characterization Plan] conceptual design for horizontal emplacement: Revision 1

    International Nuclear Information System (INIS)

    1987-08-01

    This report is a preliminary postclosure performance assessment of the repository design specified in the Site Characterization Plan Conceptual Design Report (SCP-CDR) for horizontal emplacement of high-level nuclear waste. At the time that these analyses were done, horizontal emplacement was the preferred orientation for the waste packages but vertical emplacement is now the reference design. This assessment consists of (1) a review of the regulatory requirements and strategy to demonstrate compliance with these requirements, (2) an analysis of the performance of the total repository system, (3) an analysis of the thermomechanical behavior of the repository, (4) an analysis of brine mobility in the repository, (5) an analysis of the waste package performance, (6) an analysis of the performance of seals, and (7) comments on the sensitivity of the various performance measures to uncertainties in the data and models. These are preliminary analyses and, in most cases, involve bounding calculations of the repository behavior. They have several purposes including (1) assessing how well this conceptual design ''measures up'' against requirements, (2) gaining experience in implementing the performance assessment strategy and tools and thereby learning where improvements are needed, (3) helping to identify needed data, and (4) helping to indicate required design modifications. 26 refs., 40 figs., 20 tabs

  15. The impact of thermal loading on repository performance at Yucca Mountain

    International Nuclear Information System (INIS)

    Buscheck, T.A.; Nitao, J.J.

    1992-01-01

    In the unsaturated zone at Yucca Mountain, liquid flow along preferential fracture pathways is the only credible mechanism capable of bringing water to waste packages and transporting radionuclide to the water table. Three categories of features or mechanisms will mitigate the impact of flow along preferential fracture pathways: (1) discontinuity in fracture pathways, (2) liquid-phase dispersion in fracture networks, and (3) fracture-matrix interaction. For repository areal power densities (APDs) that are too low to result in significant boiling or rock dry-out effects, the primary mode of fracture-matrix interaction is matrix imbibition. For high APDs, boiling and enhanced matrix imbibition due to rock dry-out significantly add to the capacity of the unsaturated zone to retard fracture-dominated flow. With the use of V-TOUGH code, hydrothermal flow calculations are made for a range of APDs and spent fuel ages. For APD > 20 kW/acre, repository-heat-generated flow of vapor and liquid in fractures is found to dominate the ambient hydrological system. For high APDs, boiling conditions can persist for 10,000 yr or longer and rock-dry benefits for at least 100,000 yr

  16. Probabilistic risk assessment for salt repository conceptual design of subsurface facilities: A techical basis for Q-list determination

    International Nuclear Information System (INIS)

    Chen, C.P.; Mayberry, J.J.; Shepherd, J.; Koza, H.; Rahmani, H.; Sinsky, J.

    1987-12-01

    Subpart G ''Quality Assurance'' of 10 CFR Part 60 requires that the US Department of Energy (DOE) apply a quality assurance program to ''all systems, structures, and components important to safety'' and to ''design and characterization of barriers important to waste isolation.'' In April 1986, DOE's Office of Geologic Repositories (OGR) issued general guidance for formulating a list of such systems, structures, and components---the Q-list. This guidance called for the use of probabilistic risk assessment (PRA) techniques to identify Q-list items. In this report, PRA techniques are applied to the underground facilities and systems described in the conceptual design report for the Salt Repository Project (SRP) in Deaf Smith County, Texas. Based on probability and dose consequence calculations, no specific items were identified for the Q-list. However, evaluation of the analyses indicated that two functions are important in precluding off-site releases of radioactivity: disposal container integrity; and isolation of the underground facility by the heating, ventilation, and air conditioning (HVAC) systems. Items related to these functions are recommended for further evaluation as the repository design progresses. 13 refs., 20 figs

  17. 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

    the site studies are completed, the time should be right to produce both a more comprehensive safety assessment of SFL 3-5 based on relevant site-specific data, and a more thoroughly developed repository design proposal. SKB has made a comprehensive effort to map a large number of correlation factors in order to determine the contents of relevant nuclides in the waste. However, SKI and SSI feel that the SKB choice of correlation factors could be better justified in a number of cases, particularly with respect to transuraniums and certain fission products. The authorities find that the SKB method of choosing conservative values for the present assessment was not followed for all radionuclides. Nor is the SKB safety assessment comprehensive enough to allow SKB, at this stage, to draw conclusions with the intention of refining their continued research to include only a limited selection of the relevant radionuclides. On the other hand, SKB's preliminary safety assessment has identified a number of radionuclides that definitely cannot be deprioritised from future evaluations. SKB have performed an extremely ambitious review of the waste that is intended to be deposited in SFL 3-5. The authorities consider the review of the waste to be deposited in SFL 5 to be particularly well-documented and highly traceable. The difficulties involved in accurate characterisation are greater in connection with the historic waste that is being stored at Studsvik and is intended for deposition in SFL 3. This waste is highly non-homogeneous and derives from many different sources. Moreover, a large part of this waste has already been conditioned, making any further characterisation difficult

  18. Building quality into performance and safety assessment software

    International Nuclear Information System (INIS)

    Wojciechowski, L.C.

    2011-01-01

    Quality assurance is integrated throughout the development lifecycle for performance and safety assessment software. The software used in the performance and safety assessment of a Canadian deep geological repository (DGR) follows the CSA quality assurance standard CSA-N286.7 [1], Quality Assurance of Analytical, Scientific and Design Computer Programs for Nuclear Power Plants. Quality assurance activities in this standard include tasks such as verification and inspection; however, much more is involved in producing a quality software computer program. The types of errors found with different verification methods are described. The integrated quality process ensures that defects are found and corrected as early as possible. (author)

  19. A GoldSim Model and a Sensitivity Study for Safety Assessment of a Repository for Disposal of Spent Nuclear Fuel

    International Nuclear Information System (INIS)

    Lee, Youn Myoung; Hwang, Yong Soo

    2008-11-01

    An assessment program for the evaluation of a high-level waste (HLW) repository has been developed by utilizing GoldSim, by which nuclide transports 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 could be modeled and evaluated. To demonstrate its usability, three illustrative cases including the influence of a groundwater flow pattern through canisters associated with a flowing groundwater through fractures, and the possible disruptive events caused by an accidental human intrusion or an earthquake have been investigated and illustrated for a hypothetical Korean HLW repository

  20. An assessment of the radiological impact of human intrusion at the UK Low Level Waste Repository (LLWR) - 59356

    International Nuclear Information System (INIS)

    Hicks, Tim; Baldwin, Tamara; Cummings, Richard; Sumerling, Trevor

    2012-01-01

    The UK Low Level Waste Repository Ltd submitted an Environmental Safety Case for the disposal of low-level waste (LLW) to the Environment Agency on the 1 May 2011. The Environmental Safety Case (ESC) presents a complete case for the environmental safety of the Low Level Waste Repository (LLWR) both during operations and in the long term (Cummings et al, in these proceedings). This includes an assessment of the long-term radiological safety of the facility, including an assessment of the potential consequences of human intrusion at the site. The human intrusion assessment is based on a cautiously realistic approach in defining intrusion cases and parameter values. A range of possible human intrusion events was considered based on present-day technologies and credible future uses of the site. This process resulted in the identification of geotechnical investigations, a housing development and a smallholding as requiring quantitative assessment. A particular feature of the site is that, because of its proximity to the coast and in view of expected global sea-level rise, it is vulnerable to coastal erosion. During such erosion, wastes and engineered barrier materials will be exposed, and could become targets for investigation or recovery. Therefore, human intrusion events have been included that are associated with such activities. A radiological assessment model has been developed to analyse the impacts of potential human intrusion at the site. A key feature of the model is the representation of the spatial layout of the disposal site, including the engineered cap design and the large-scale spatial heterogeneity of radionuclide concentrations within the repository. The model has been used to calculate the radiation dose to intruders and to others following intrusion at different times and at different locations across the site, for the each of the selected intrusion events, considering all relevant exposure modes. Potential doses due to radon and its daughters in

  1. Assessment of crushed salt consolidation and fracture healing processes in a nuclear waste repository in salt

    International Nuclear Information System (INIS)

    1984-11-01

    For a nuclear waste repository in salt, two aspects of salt behavior are expected to contribute to favorable conditions for waste isolation. First, consolidation of crushed salt backfill due to creep closure of the underground openings may result in a backfill barrier with low permeability. Second, fractures created in the salt by excavation may heal under the influence of stress and temperature following sealing. This report reviews the status of knowledge regarding crushed salt consolidation and fracture healing, provides analyses which predict the rates at which the processes will occur under repository conditions, and develops requirements for future study. Analyses of the rate at which crushed salt will consolidate are found to be uncertain because of unexplained wide variation in the creep properties of crushed salt obtained from laboratory testing, and because of uncertainties in predictions of long term closure rates of openings in salt. This uncertainty could be resolved to a large degree by additional laboratory testing of crushed salt. Similarly, additional testing of fracture healing processes is required to confirm that healing will be effective under repository conditions. Extensive references, 27 figures, 5 tables

  2. The impact of thermal loading on repository performance at Yucca Mountain

    International Nuclear Information System (INIS)

    Buscheck, T.A.; Nitao, J.J.

    1992-01-01

    This paper reports that in the unsaturated zone at Yucca Mountain, liquid flow along preferential fracture pathways is the only credible mechanism capable of bringing water to waste packages and transporting radionuclides to the water table. Three categories of features or mechanisms will mitigate the impact of flow along preferential fracture pathways: discontinuity in fracture pathways, liquid-phase dispersion in fracture networks, and fracture-matrix interaction. For repository areal power densities (APDs) that are too low to result in significant boiling or rock dry-out effects, the primary mode of fracture-matrix interaction is matrix imbibition. For high APDs, boiling and enhanced matrix imbibition due to rock dry-out significantly add to the capacity of the unsaturated zone to retard fracture-dominated flow

  3. OLEM Performance Assessment Information

    Data.gov (United States)

    U.S. Environmental Protection Agency — This asset includes a variety of data sets that measure the performance of Office of Land and Emergency Management (OLEM) programs in support of the Office of the...

  4. Simply Performance Assessment

    Science.gov (United States)

    McLaughlin, Cheryl A.; McLaughlin, Felecia C.; Pringle, Rose M.

    2013-01-01

    This article presents the experiences of Miss Felecia McLaughlin, a fourth-grade teacher from the island of Jamaica who used the model proposed by Bass et al. (2009) to assess conceptual understanding of four of the six types of simple machines while encouraging collaboration through the creation of learning teams. Students had an opportunity to…

  5. 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

  6. THE PENA BLANCA NATURAL ANALOGUE PERFORMANCE ASSESSMENT MODEL

    Energy Technology Data Exchange (ETDEWEB)

    G. Saulnier and W. Statham

    2006-04-16

    The Nopal I uranium mine in the Sierra Pena Blanca, Chihuahua, Mexico serves as a natural analogue to the Yucca Mountain repository. The Pena Blanca Natural Analogue Performance Assessment Model simulates the mobilization and transport of radionuclides that are released from the mine and transported to the saturated zone. The Pena Blanca Natural Analogue Performance Assessment Model uses probabilistic simulations of hydrogeologic processes that are analogous to the processes that occur at the Yucca Mountain site. The Nopal I uranium deposit lies in fractured, welded, and altered rhyolitic ash-flow tuffs that overlie carbonate rocks, a setting analogous to the geologic formations at the Yucca Mountain site. The Nopal I mine site has the following analogous characteristics as compared to the Yucca Mountain repository site: (1) Analogous source--UO{sub 2} uranium ore deposit = spent nuclear fuel in the repository; (2) Analogous geology--(i.e. fractured, welded, and altered rhyolitic ash-flow tuffs); (3) Analogous climate--Semiarid to arid; (4) Analogous setting--Volcanic tuffs overlie carbonate rocks; and (5) Analogous geochemistry--Oxidizing conditions Analogous hydrogeology: The ore deposit lies in the unsaturated zone above the water table.

  7. THE PENA BLANCA NATURAL ANALOGUE PERFORMANCE ASSESSMENT MODEL

    International Nuclear Information System (INIS)

    G. Saulnier; W. Statham

    2006-01-01

    The Nopal I uranium mine in the Sierra Pena Blanca, Chihuahua, Mexico serves as a natural analogue to the Yucca Mountain repository. The Pena Blanca Natural Analogue Performance Assessment Model simulates the mobilization and transport of radionuclides that are released from the mine and transported to the saturated zone. The Pena Blanca Natural Analogue Performance Assessment Model uses probabilistic simulations of hydrogeologic processes that are analogous to the processes that occur at the Yucca Mountain site. The Nopal I uranium deposit lies in fractured, welded, and altered rhyolitic ash-flow tuffs that overlie carbonate rocks, a setting analogous to the geologic formations at the Yucca Mountain site. The Nopal I mine site has the following analogous characteristics as compared to the Yucca Mountain repository site: (1) Analogous source--UO 2 uranium ore deposit = spent nuclear fuel in the repository; (2) Analogous geology--(i.e. fractured, welded, and altered rhyolitic ash-flow tuffs); (3) Analogous climate--Semiarid to arid; (4) Analogous setting--Volcanic tuffs overlie carbonate rocks; and (5) Analogous geochemistry--Oxidizing conditions Analogous hydrogeology: The ore deposit lies in the unsaturated zone above the water table

  8. Consideration of environmental change in performance assessments

    International Nuclear Information System (INIS)

    Pinedo, P.; Thorne, M.; Egan, M.; Calvez, M.; Kautsky, U.

    2005-01-01

    Depending on the particular circumstances in which a post-closure performance assessment of a radioactive waste repository is made, it may be appropriate to follow simple or more complex approaches in characterising the biosphere. Several different Example Reference Biospheres were explored in BIOMASS Theme 1 to address a range of issues that arise. Here, consideration is given to Example Reference Biospheres relevant to representing the implications of changes that may occur within the biosphere system during the period over which releases of radionuclides from a disposal facility might take place. Mechanisms of change considered include those extrinsic and intrinsic to the system of interest. An overall methodology for incorporating environmental change into assessments is proposed. This includes screening of primary mechanisms of change; identification of possible time sequences of change; development of a coherent description of the regional landscape response for each time sequence; integration of source term and geosphere-biosphere interface information; identification and description of one or more time series of assessment biospheres; and evaluation of the advantages and disadvantages of simulating the effects of sequences of biosphere systems and the transitions between them, or of defining a set of biosphere systems to be represented individually in a non-sequential analysis. The usefulness of the methodology is explored in two site-specific examples and one generic example

  9. Total System Performance Assessment

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Yong Soo; Kang, Chul Hyung; Lee, Youn Myoung; Han, Ji Woong; Choi, Jong Won; Hahn, Pil Soo; Park, Jeong Hwa; Jeong, Mi Seon

    2007-06-15

    Based on the KAERI FEP list developed through the previous studies, the KAERI FEP Encyclopedia has been developed. Current version is 1.0 which includes all relevant FEPs to compose of two references and all alternative scenarios. Many interaction FEPs between scenario defining FEP(SDF) are created throughout the study. FEPs are classified into many Integrated FEP(IFEP) which eventually become the elements of the RES matrix. The FEAS program one of the component of the KAERI's CYPRUS information system is added to develop the FEP, RES, AC, AMF and finally scenarios. It assists to create transparent way to deal with assessment from the stage of the planning of the R and D to the final stage of the external audit and regulatory body review. Even though MASCOT-K and compartment analysis codes such as AMBER, GoldSim and Ecolego are excellent for TSPA they by in heritage possess a certain limitation especially to identify a proper migration cross sectional area when a relatively big component intersects with a tiny one such as a fracture. It is truly 3D phenomena in nature. MDPSA code is developed which is expected to overcome limitations in compartment models while successfully deals with natural disruptive events. The R and D target for the TSPA is to develop the sufficient scenarios and their variation cases to understand the safety of KRS in every possible aspect. For this, reference scenarios, alternative scenarios covering engineered barrier failure and natural events are developed and assessed respectively for around 100 cases. The stylized template to assess the Korean reference biosphere is developed using the AMBER. Three critical groups, agricultural, freshwater and marine water fishing groups are identified to assess the DCF following the guidelines of ICRP. Based on the QA principles of T2R3, the web based QA system is developed using the procedures in the USNRC 10CFR50 Appendix B. The QA system is combined with the PAID and FEAS to create the comprehensive

  10. Total System Performance Assessment

    International Nuclear Information System (INIS)

    Hwang, Yong Soo; Kang, Chul Hyung; Lee, Youn Myoung; Han, Ji Woong; Choi, Jong Won; Hahn, Pil Soo; Park, Jeong Hwa; Jeong, Mi Seon

    2007-06-01

    Based on the KAERI FEP list developed through the previous studies, the KAERI FEP Encyclopedia has been developed. Current version is 1.0 which includes all relevant FEPs to compose of two references and all alternative scenarios. Many interaction FEPs between scenario defining FEP(SDF) are created throughout the study. FEPs are classified into many Integrated FEP(IFEP) which eventually become the elements of the RES matrix. The FEAS program one of the component of the KAERI's CYPRUS information system is added to develop the FEP, RES, AC, AMF and finally scenarios. It assists to create transparent way to deal with assessment from the stage of the planning of the R and D to the final stage of the external audit and regulatory body review. Even though MASCOT-K and compartment analysis codes such as AMBER, GoldSim and Ecolego are excellent for TSPA they by in heritage possess a certain limitation especially to identify a proper migration cross sectional area when a relatively big component intersects with a tiny one such as a fracture. It is truly 3D phenomena in nature. MDPSA code is developed which is expected to overcome limitations in compartment models while successfully deals with natural disruptive events. The R and D target for the TSPA is to develop the sufficient scenarios and their variation cases to understand the safety of KRS in every possible aspect. For this, reference scenarios, alternative scenarios covering engineered barrier failure and natural events are developed and assessed respectively for around 100 cases. The stylized template to assess the Korean reference biosphere is developed using the AMBER. Three critical groups, agricultural, freshwater and marine water fishing groups are identified to assess the DCF following the guidelines of ICRP. Based on the QA principles of T2R3, the web based QA system is developed using the procedures in the USNRC 10CFR50 Appendix B. The QA system is combined with the PAID and FEAS to create the comprehensive

  11. Total System Performance Assessment

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Yong Soo; Kang, Chul Hyung; Lee, Youn Myoung; Han, Ji Woong; Choi, Jong Won; Hahn, Pil Soo; Park, Jeong Hwa; Jeong, Mi Seon

    2007-06-15

    Based on the KAERI FEP list developed through the previous studies, the KAERI FEP Encyclopedia has been developed. Current version is 1.0 which includes all relevant FEPs to compose of two references and all alternative scenarios. Many interaction FEPs between scenario defining FEP(SDF) are created throughout the study. FEPs are classified into many Integrated FEP(IFEP) which eventually become the elements of the RES matrix. The FEAS program one of the component of the KAERI's CYPRUS information system is added to develop the FEP, RES, AC, AMF and finally scenarios. It assists to create transparent way to deal with assessment from the stage of the planning of the R and D to the final stage of the external audit and regulatory body review. Even though MASCOT-K and compartment analysis codes such as AMBER, GoldSim and Ecolego are excellent for TSPA they by in heritage possess a certain limitation especially to identify a proper migration cross sectional area when a relatively big component intersects with a tiny one such as a fracture. It is truly 3D phenomena in nature. MDPSA code is developed which is expected to overcome limitations in compartment models while successfully deals with natural disruptive events. The R and D target for the TSPA is to develop the sufficient scenarios and their variation cases to understand the safety of KRS in every possible aspect. For this, reference scenarios, alternative scenarios covering engineered barrier failure and natural events are developed and assessed respectively for around 100 cases. The stylized template to assess the Korean reference biosphere is developed using the AMBER. Three critical groups, agricultural, freshwater and marine water fishing groups are identified to assess the DCF following the guidelines of ICRP. Based on the QA principles of T2R3, the web based QA system is developed using the procedures in the USNRC 10CFR50 Appendix B. The QA system is combined with the PAID and FEAS to create the