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Sample records for safety analysis sfr

  1. SFR Safety Considerations

    International Nuclear Information System (INIS)

    Glatz, Jean-Paul

    2012-01-01

    Objectives of the Safety and Operation Project: • analysis and experiments that support approaches and assess performance of specific safety features, • development and verification of computational tools and validation of models employed in safety assessment and facility licensing, and • valorisation of reactor operation, from experience and testing in operating SFR plants

  2. Safety analysis SFR 1. Long-term safety

    Energy Technology Data Exchange (ETDEWEB)

    2008-12-15

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

  3. Safety analysis SFR 1. Long-term safety

    International Nuclear Information System (INIS)

    2008-12-01

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

  4. Safety analysis SFR 1. Long-term safety

    Energy Technology Data Exchange (ETDEWEB)

    2008-12-15

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

  5. Review of SFR Design Safety using Preliminary Regulatory PSA Model

    International Nuclear Information System (INIS)

    Na, Hyun Ju; Lee, Yong Suk; Shin, Andong; Suh, Nam Duk

    2013-01-01

    The major objective of this research is to develop a risk model for regulatory verification of the SFR design, and thereby, make sure that the SFR design is adequate from a risk perspective. In this paper, the development result of preliminary regulatory PSA model of SFR is discussed. In this paper, development and quantification result of preliminary regulatory PSA model of SFR is discussed. It was confirmed that the importance PDRC and ADRC dampers is significant as stated in the result of KAERI PSA model. However, the importance can be changed significantly depending on assumption of CCCG and CCF factor of PDRC and ADRC dampers. SFR (sodium-cooled fast reactor) which is Gen-IV nuclear energy system, is designed to accord with the concept of stability, sustainability and proliferation resistance. KALIMER-600, which is under development in Korea, includes passive safety systems (e. g. passive reactor shutdown, passive residual heat removal, and etc.) as well as active safety systems. Risk analysis from a regulatory perspective is needed to support the regulatory body in its safety and licensing review for SFR (KALIMER-600). Safety issues should be identified in the early design phase in order to prevent the unexpected cost increase and delay of the SFR licensing schedule that may be caused otherwise

  6. Development of Basic Key Technologies for Gen IV SFR Safety Evaluation

    International Nuclear Information System (INIS)

    Jeong, Hae Yong; Kwon, Young Min; Kim, Tae Woon; Park, Soo Yong; Suk, Soo Dong; Lee, Kwi Lim; Lee, Yong Bum; Chang, Won Pyo; Ha, Kwi Seok; Hahn, Sang Hoon

    2010-07-01

    Safety issues and design requirements on control rod worth were identified through the evaluation of safety design characteristics and the preliminary safety evaluation. This results will be taken into account for the conceptual design studies of the demonstration reactor in the next stage. The Level-1 Pasa has been performed and a quantitative Cdf value was produced for the selected design from the several candidates. The inherent safety characteristics of the selected design were evaluated through the DBE and ATWS analyses. A surrogate material for Tru has been selected which is applicable to the study of liquidus/solidus temperature test for the metallic fuel containing Tru. A methodology for the regression analysis with surrogate material has been developed and valuable data on metal fuel liquidus/solidus temperature have been measured. A simple mechanistic model describing a bending of subassemblies has been formulated based on the foreign test data and existing models. Its applicability has been evaluated for the Phenix design. New criteria of the core damage for the SFR PSA were identified. The list of initiating events, system response event tree, and core response event tree, which constitute a PSA methodology for an SFR, have been introduced. By developing the SFR PIRT, phenomenological model features, which have to be satisfied in a safety code, were defined and the PIRT results were applied to the design of the PDRC test facility. Bases for a safety evaluation methodology for the SFR DBEs have been also prepared. A draft version of the topical report on the code for local fault analysis has been completed. Since 2007, the MARS-LMR code has been developed and assessments for model validation with the test data from EBR-II and Phenix reactor have been continued. The code has been applied to the evaluation of passive safety of a conceptual design of Gen IV SFR

  7. An Innovative Hybrid Loop-Pool SFR Design and Safety Analysis Methods: Today and Tomorrow

    International Nuclear Information System (INIS)

    Hongbin Zhang; Haihua Zhao; Vincent Mousseau

    2008-01-01

    Investment in commercial sodium cooled fast reactor (SFR) power plants will become possible only if SFRs achieve economic competitiveness as compared to light water reactors and other Generation IV reactors. Toward that end, we have launched efforts to improve the economics and safety of SFRs from the thermal design and safety analyses perspectives at Idaho National Laboratory. From the thermal design perspective, an innovative hybrid loop-pool SFR design has been proposed. This design takes advantage of the inherent safety of a pool design and the compactness of a loop design to further improve economics and safety. From the safety analyses perspective, we have initiated an effort to develop a high fidelity reactor system safety code

  8. Project SAFE. Update of the SFR-1 safety assessment. Phase 1

    International Nuclear Information System (INIS)

    Andersson, Johan; Riggare, P.; Skagius, K.

    1998-10-01

    SFR-1 is a facility for disposal of low-level radioactive operational waste from the nuclear power plants in Sweden. Low-level radioactive waste from industry, medicine, and research is also disposed in SFR-1. The facility is situated in bedrock beneath the Baltic Sea, 1 km off the coast near the Forsmark nuclear power plant. SFR-1 was built between the years 1983 and 1988. An assessment of the long-term performance of the facility was included in the vast documentation that was a part of the application for an operational license. The assessment was presented in the form of a final safety report. In the operational licence for SFR-1 it is stated that renewed safety assessments should be carried out at least each ten years. In order to meet this demand SKB has launched a special project, SAFE (Safety Assessment of Final Disposal of Operational Radioactive Waste). The aim of the project is to update the safety analysis and to prepare a safety report that will be presented to the Swedish authorities not later than year 2000. Project SAFE is divided into three phases. The first phase is a prestudy, and the results of the prestudy are given in this report. The aim of the prestudy is to identify issues where additional studies would improve the basis for the updated safety analysis as well as to suggest how these studies should be carried out. The work has been divided into six different topics, namely the inventory, the near field, the far field, the biosphere, radionuclide transport calculations and scenarios. For each topic the former safety reports and regulatory reviews are scrutinised and needs for additional work is identified. The evaluations are given in appendices covering the respective topics. The main report is a summary of the appendices with a more stringent description of the repository system and the processes that are of interest and therefore should be addressed in an updated safety assessment. However, it should be pointed out that one of the

  9. Improving SFR Economics through Innovations from Thermal Design and Analysis Aspects

    Energy Technology Data Exchange (ETDEWEB)

    Haihua Zhao; Hongbin Zhang; Vincent Mousseau; Per F. Peterson

    2008-06-01

    Achieving economic competitiveness as compared to LWRs and other Generation IV (Gen-IV) reactors is one of the major requirements for large-scale investment in commercial sodium cooled fast reactor (SFR) power plants. Advances in R&D for advanced SFR fuel and structural materials provide key long-term opportunities to improve SFR economics. In addition, other new opportunities are emerging to further improve SFR economics. This paper provides an overview on potential ideas from the perspective of thermal hydraulics to improve SFR economics. These include a new hybrid loop-pool reactor design to further optimize economics, safety, and reliability of SFRs with more flexibility, a multiple reheat and intercooling helium Brayton cycle to improve plant thermal efficiency and reduce safety related overnight and operation costs, and modern multi-physics thermal analysis methods to reduce analysis uncertainties and associated requirements for over-conservatism in reactor design. This paper reviews advances in all three of these areas and their potential beneficial impacts on SFR economics.

  10. Assessment of the long-term safety for SFR

    Energy Technology Data Exchange (ETDEWEB)

    Greis Dahlberg, Christina; Vahlund, Frederik [Svensk Kaernbraenslehantering AB, Stockholm (Sweden)

    2015-07-01

    During operation and decommissioning of the Swedish nuclear facilities, radioactive waste is generated that must be disposed of. Besides waste from the nuclear facilities, some waste derives from other activities such as industry, research, medical care, etc. Short-lived low- and intermediate-level waste from these activities is disposed of in the final repository for short-lived radioactive waste, SFR, in Forsmark. The facility, which has been in operation since 1988, is owned and operated by Svensk Karnbranslehantering AB, SKB. The existing facility has neither sufficient space nor a license to receive decommissioning waste. SFR must therefore be extended so that shortlived low- and intermediate-level decommissioning waste from the nuclear facilities can also be received. The need for additional capacity has been accentuated by the closure of two reactors in Barseback. These reactors cannot be dismantled until the SFR facility has been extended. The existing repository is built to receive, and after closure serve as a passive repository for, low- and intermediate-level radioactive waste. The disposal rooms are situated in the bedrock beneath the sea floor, covered by about 60 metres of rock. The repository has been designed so that it can be abandoned after closure without requiring further measures to maintain its function. The extension of SFR, is done at the -120 m level immediately adjacent to, and within the same depth range as, the existing facility. The basic function of the existing SFR and of the extended one will be the same. However, a clear difference is the design of the tunnel and the rock vault that are required to permit transport and storage of whole reactor pressure vessels. The application for a license to build this extension includes an assessment of the long-term safety (post-closure safety) of the facility. The safety assessment also contains an updated assessment of the long-term safety of the existing facility. The safety assessment for

  11. Development Perspective of Regulatory Audit Code System for SFR Nuclear Safety Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Moo Hoon; Lee, Gil Soo; Shin, An Dong; Suh, Nam Duk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2012-05-15

    A sodium-cooled fast reactor (SFR) in Korea is based on the KALIMER-600 concept developed by KAERI. Based on 'Long-term R and D Plan for Future Reactor Systems' which was approved by the Korea Atomic Energy Commission in 2008, the KAERI designer is scheduled to apply the design certification of the prototype SFR in 2017. In order to establish regulatory infrastructure for the licensing of a prototype SFR, KINS has develop the regulatory requirements for the demonstration SFR since 2010, and are scheduled to develop the regulatory audit code systems in regard to core, fuel, and system, etc. since 2012. In this study, the domestic code systems used for core design and safety evaluation of PWRs and the nuclear physics and code system for SFRs were briefly reviewed, and the development perspective of regulatory audit code system for SFR nuclear safety evaluation were derived

  12. Safety design approach for JSFR toward the realization of GEN IV SFR

    International Nuclear Information System (INIS)

    Kubo, S.; Yamano, H.; Chikazawa, Y.; Shimakawa, Y.

    2013-01-01

    Conclusion: Safety Design Approach for JSFR: • Based on the safety design criteria for Generation-IV SFR • DECs, Situations practically eliminated and related design measures are identified and selected with due consideration of the safety features of SFR and the lessons learned from the TEPCO’s Fukushima Dai-ichi nuclear power plants accident Safety Design Concept of JSFR: • For failure to shutdown: Passive shutdown capability, Mitigation of core damage (Prevention of severe mechanical energy release, In-Vessel Retention) • For failure to remove heat: Prevention of significant core damage (Natural circulation DHR, Alternative cooling measures) • Containment: Prevention of sever dynamic loads by design measures (IVR, double boundary concept, inertization)

  13. Analysis of SCARABEE BE+3 experiment with ASTEC-Na and comparison with other SFR safety analysis codes

    International Nuclear Information System (INIS)

    Bandini, Giacomino; Ederli, Stefano; Perez-Martin, Sara; Pfrang, Werner; Girault, Nathalie; Cloarec, Laure

    2017-01-01

    The ASTEC-Na code was further developed and assessed in the frame of JASMIN project of the 7th EU Framework Program to extend the original capability of ASTEC, dealing with severe accident analysis in LWR to Sodium-cooled Fast Reactors (SFR). The in-pile BE+3 experiment from the SCARABEE-N program has been simulated with ASTEC-Na for thermal-hydraulic models validation purpose. The adequacy of ASTEC-Na thermal-hydraulic models has been also investigated through the comparison with other safety analysis codes. The analysis of SCARABEE BE+3 test confirms the good performance of ASTEC-Na code in the calculation of single-phase conditions and boiling onset, while larger deviations are encountered in the analysis of the two-phase conditions, mainly regarding the propagation of the boiling front. Furthermore, reasonable agreement was found with other code results. (author)

  14. R&D Challenges for SFR Design and Safety Analysis – Opportunities for International Cooperations

    International Nuclear Information System (INIS)

    Devictor, Nicolas

    2013-01-01

    Examples of R&D challenges related to safety have been presented. For any domain, R&D activities includes modelling, codes development and their V&V process, with the support of experimental programs. The success in the R&D will help the safety case and the acceptability of SFR. Some of these activities are relevant for international cooperation especially benchmarks and sharing of experimental facilities. This last point could take benefit of recent catalogues experimental facilities (already operational or in project), for example from the TAREF Task Force of OECD/NEA and the European project ADRIANA

  15. SSI and SKI's Review of SKB's Updated Final Safety Report for SFR 1. Review Report

    International Nuclear Information System (INIS)

    2003-10-01

    previous safety assessments for SFR 1. The review committee particularly emphasises the high quality of the site-specific description prepared for the biosphere and biosphere evolution. Significant progress has also been made with respect to the basis for analyses of the inventory of radioactive substances and model descriptions for hydrogeology and near field transport. The safety assessment has also been updated with a modern system description and database for FEPs (features, events and processes). The requirements made on a repository and on the content and presentation of a safety report have been stipulated in the form of new regulations since permission was granted for the operation of SFR 1. The regulations stipulate that the consequences of radioactive releases are to be evaluated in relation to a risk criterion and that environmental impact must also be reported. In connection with the evaluation of SKB's new safety assessment, the authorities identified certain deviations in relation to the requirements of the new regulations. A significant deviation is that SSR 2001 lacks a thorough analysis which shows the impact of identified uncertainties on calculated future consequences. There are also deficiencies in the documentation which make it difficult to reproduce or judge the suitability of certain models and data, e.g. for the near field and biosphere. Confidence in SKB's risk assessment is also limited by the lack of clarity of the scenario descriptions. Furthermore, SKB has not convincingly shown that the assessment can be limited to 10,000 years. In the opinion of the review committee, the radionuclide inventory that SKB refers to as conservative, and which has been used as an assumption in the calculations, is beyond the limits of the present licence. In the case of the realistic inventory, there is adequate agreement with the licensed inventory. It cannot be excluded that the inventory of certain long-lived and significant radionuclides may be underestimated

  16. RISK-INFORMED BALANCING OF SAFETY, NONPROLIFERATION, AND ECONOMICS FOR THE SFR

    Energy Technology Data Exchange (ETDEWEB)

    Apostolakis, George; Driscoll, Michael; Golay, Michael; Kadak, Andrew; Todreas, Neil; Aldmir, Tunc; Denning, Richard; Lineberry, Michael

    2011-10-20

    A substantial barrier to the implementation of Sodium-cooled Fast Reactor (SFR) technology in the short term is the perception that they would not be economically competitive with advanced light water reactors. With increased acceptance of risk-informed regulation, the opportunity exists to reduce the costs of a nuclear power plant at the design stage without applying excessive conservatism that is not needed in treating low risk events. In the report, NUREG-1860, the U.S. Nuclear Regulatory Commission describes developmental activities associated with a risk-informed, scenario-based technology neutral framework (TNF) for regulation. It provides quantitative yardsticks against which the adequacy of safety risks can be judged. We extend these concepts to treatment of proliferation risks. The objective of our project is to develop a risk-informed design process for minimizing the cost of electricity generation within constraints of adequate safety and proliferation risks. This report describes the design and use of this design optimization process within the context of reducing the capital cost and levelized cost of electricity production for a small (possibly modular) SFR. Our project provides not only an evaluation of the feasibility of a risk-informed design process but also a practical test of the applicability of the TNF to an actual advanced, non-LWR design. The report provides results of five safety related and one proliferation related case studies of innovative design alternatives. Applied to previously proposed SFR nuclear energy system concepts We find that the TNF provides a feasible initial basis for licensing new reactors. However, it is incomplete. We recommend improvements in terms of requiring acceptance standards for total safety risks, and we propose a framework for regulation of proliferation risks. We also demonstrate methods for evaluation of proliferation risks. We also suggest revisions to scenario-specific safety risk acceptance standards

  17. Risk-Informed Balancing Of Safety, Nonproliferation, And Economics For The SFR

    International Nuclear Information System (INIS)

    Apostolakis, George; Driscoll, Michael; Golay, Michael; Kadak, Andrew; Todreas, Neil; Aldmir, Tunc; Denning, Richard; Lineberry, Michael

    2011-01-01

    A substantial barrier to the implementation of Sodium-cooled Fast Reactor (SFR) technology in the short term is the perception that they would not be economically competitive with advanced light water reactors. With increased acceptance of risk-informed regulation, the opportunity exists to reduce the costs of a nuclear power plant at the design stage without applying excessive conservatism that is not needed in treating low risk events. In the report, NUREG-1860, the U.S. Nuclear Regulatory Commission describes developmental activities associated with a risk-informed, scenario-based technology neutral framework (TNF) for regulation. It provides quantitative yardsticks against which the adequacy of safety risks can be judged. We extend these concepts to treatment of proliferation risks. The objective of our project is to develop a risk-informed design process for minimizing the cost of electricity generation within constraints of adequate safety and proliferation risks. This report describes the design and use of this design optimization process within the context of reducing the capital cost and levelized cost of electricity production for a small (possibly modular) SFR. Our project provides not only an evaluation of the feasibility of a risk-informed design process but also a practical test of the applicability of the TNF to an actual advanced, non-LWR design. The report provides results of five safety related and one proliferation related case studies of innovative design alternatives. Applied to previously proposed SFR nuclear energy system concepts We find that the TNF provides a feasible initial basis for licensing new reactors. However, it is incomplete. We recommend improvements in terms of requiring acceptance standards for total safety risks, and we propose a framework for regulation of proliferation risks. We also demonstrate methods for evaluation of proliferation risks. We also suggest revisions to scenario-specific safety risk acceptance standards

  18. SFR 1 Vault Database

    International Nuclear Information System (INIS)

    Savage, David; Stenhouse, Mike

    2002-04-01

    SKB is carrying out a safety assessment of the operational SFR 1 repository under the auspices of the 'SAFE' (Safety Assessment of Final Repository for Radioactive Operational Waste) project. SKI in turn, is carrying out its own review of SFR 1. The work presented here is a compilation of physical and chemical data for the SFR 1 repository which will be used in radionuclide transport and assessment calculations by SKI. This compilation has focused on the repository itself (engineered barriers plus near-field rock). Data have been compiled for the following: Physical properties (porosity, hydraulic conductivity, bulk density, effective diffusivity); Sorption of radionuclides (on concrete, sand, bentonite, sand-bentonite, and rock); Radionuclide solubility. In addition, issues affecting gas generation at SFR I have been reviewed and placed in context with research conducted for the SFL 3-5 repository

  19. Preliminary Development of Regulatory PSA Models for SFR

    International Nuclear Information System (INIS)

    Choi, Yong Won; Shin, Andong; Bae, Moohoon; Suh, Namduk; Lee, Yong Suk

    2013-01-01

    Well developed PRA methodology exists for LWR (Light Water Reactor) and PHWR (Pressurized Heavy Water Reactor). Since KAERI is developing a prototype SFR targeting to apply for a license by 2017, KINS needs to have a PRA models to assess the safety of this prototype reactor. The purpose of this study is to develop the regulatory PSA models for the independent verification of the SFR safety. Since the design of the prototype SFR is not mature yet, we have tried to develop the preliminary models based on the design data of KAERI's previous SFR design. In this study, the preliminary initiating events of level 1 internal event for SFR were selected through reviews of existing PRA (LWR, PRISM, ASTRID and KALIMER-600) models. Then, the event tree for each selected initiating event was developed. The regulatory PRA models of SFR developed are preliminary in a sense, because the prototype SFR design is not mature and provided yet. Still it might be utilized for the forthcoming licensing review in assessing the risk of safety issues and the configuration control of the design

  20. Perspective on the audit calculation for SFR using TRACE code

    Energy Technology Data Exchange (ETDEWEB)

    Shin, An Dong; Choi, Yong Won; Bang, Young Suk; Bae, Moo Hoon; Huh, Byung Gil; Seol, Kwang One [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2012-10-15

    Korean Sodium Cooled Fast Reactor (SFR) is being developed by KAERI. The Prototype SFR will be a first SFR applied for licensing. KINS started research programs for preparing new concept design licensing recently. Safety analysis for the certain reactor is based on the computational estimation with conservatism and/or uncertainty of modeling. For the audit calculation for sodium cooled fast reactor (SFR), TRACE code is considered as one of analytical tool for SFR since TRACE code have already sodium related properties and models in it and have experience in the liquid metal coolant system area in abroad. Applicability of TRACE code for SFR is prechecked before real audit calculation. In this study, Demonstration Fast Reactor (DFR) 600 steady state conditions is simulated for identification of area of modeling improvements of TRACE code.

  1. Development of mechanical analysis module for simulation of SFR fuel rod behavior using finite element method

    International Nuclear Information System (INIS)

    Shin, Andong; Jeong, Hyedong; Suh, Namduk; Kim, Hyochan; Yang, Yongsik

    2014-01-01

    Korean SFR developer decided to adapt metal fuel, current study focused on the metal fuel instead of oxide fuel. The SFR metal fuel has been developed by Korea Atomic Energy Research Institute (KAERI) and many efforts focused on designing and manufacturing the metal fuel. Since a nuclear fuel is the first barrier to protect radioactive isotope release, the fuel's integrity must be secured during steady-state operation and accident condition within an acceptable range. Whereas the design and evaluation methodologies, code systems and test procedures of a light water reactor fuel are sufficiently established, those of the SFR fuel needs more technical advances. In the view of regulatory point, there are still many challenging issues which are required to secure the safety of fuel and reactors. For this reason, the Korea Institute of Nuclear Safety (KINS) has launched the new project to develop the regulatory technology for SFR system including a fuel area. The ALFUS code was developed by CRIEPI and employs mechanistic model for fission gas release and swelling of fuel slug. In the code system, a finite element method was introduced to analyze the fuel and cladding's mechanical behaviors. The FEAST code is more advanced code system for SFR which adopted mechanistic FGR and swelling model but still use analytical model to simulate fuel and cladding mechanical behavior. Based on the survey of the previous studies, fuel and cladding mechanical model should be improved. Analysis of mechanical behavior for fuel rod is crucial to evaluate overall rod's integrity. In addition, it is because contact between fuel slug and cladding or an over-pressure of rod internal pressure can cause rod failure during steady-state and other operation condition. The most of reference codes have simplified mechanical analysis model, so called 'analytical mode', because the detailed mechanical analysis requires large amount of calculation time and computing power. Even

  2. Development of mechanical analysis module for simulation of SFR fuel rod behavior using finite element method

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Andong; Jeong, Hyedong; Suh, Namduk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of); Kim, Hyochan; Yang, Yongsik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    Korean SFR developer decided to adapt metal fuel, current study focused on the metal fuel instead of oxide fuel. The SFR metal fuel has been developed by Korea Atomic Energy Research Institute (KAERI) and many efforts focused on designing and manufacturing the metal fuel. Since a nuclear fuel is the first barrier to protect radioactive isotope release, the fuel's integrity must be secured during steady-state operation and accident condition within an acceptable range. Whereas the design and evaluation methodologies, code systems and test procedures of a light water reactor fuel are sufficiently established, those of the SFR fuel needs more technical advances. In the view of regulatory point, there are still many challenging issues which are required to secure the safety of fuel and reactors. For this reason, the Korea Institute of Nuclear Safety (KINS) has launched the new project to develop the regulatory technology for SFR system including a fuel area. The ALFUS code was developed by CRIEPI and employs mechanistic model for fission gas release and swelling of fuel slug. In the code system, a finite element method was introduced to analyze the fuel and cladding's mechanical behaviors. The FEAST code is more advanced code system for SFR which adopted mechanistic FGR and swelling model but still use analytical model to simulate fuel and cladding mechanical behavior. Based on the survey of the previous studies, fuel and cladding mechanical model should be improved. Analysis of mechanical behavior for fuel rod is crucial to evaluate overall rod's integrity. In addition, it is because contact between fuel slug and cladding or an over-pressure of rod internal pressure can cause rod failure during steady-state and other operation condition. The most of reference codes have simplified mechanical analysis model, so called 'analytical mode', because the detailed mechanical analysis requires large amount of calculation time and computing power. Even

  3. Economic Analysis of Pyro-SFR Fuel Cycle

    International Nuclear Information System (INIS)

    Gao, Fanxing; Park, Byungheung; Kwon, Eunha; Ko, Wonil

    2010-01-01

    In this study, based on the material flow the economics of Pyro-SFR has been estimated. These are mainly two methodologies to perform nuclear fuel cycle cost study which is based on the material flow calculations. One is equilibrium model and the other is dynamic model. Equilibrium model focus on the batch study with the assumptions that the whole system is in a steady state and mass flow as well as the electricity production all through the fuel cycle is in equilibrium state, which calculates the electricity production within a certain period and associated material flow with reference to unit cost in order to obtain the cost of electricity. Dynamic model takes the time factor into consideration to simulate the actual cases. Compared with the dynamic analysis model, the outcome of equilibrium model is more theoretical comparisons, especially with regard to the large uncertainty of the development of the pyro-technology evaluated. In this study equilibrium model was built to calculate the material flow on a batch basis. With the unit cost being determined, the cost of each step of fuel cycle could be obtained, so does the FMC. Due to the unavoidable uncertainty with certain unit costs, evaluated cost range and uncertainty study are applied. Sensitivity analysis has also been performed to obtain the breakeven uranium price for Pyro-SFR against PWR-O T. Economics of Pyro-SFR fuel cycle scenario has been calculated and compared by employing equilibrium model. The LFCC were obtained, Pyro-SFR 7.68 mills/kWh. The Uranium price is the dominant driver of LFCC. Pyro-techniques also weight considerably in Pyro-SFR scenario. On consideration of the current unavoidable uncertainties introduced by certain cost data, cost range and triangle techniques were used to perform the uncertainty study which indicates that the gap between Pyro-SFR and PWR-O T fuel cycle scenario is relatively small

  4. Experimental Setup with Transient Behavior of Fuel Cladding of SFR

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Sang Hun; Kim, Jun Hwan; Kim, June-Hyung; Ryu, Woo Seog; Park, Sang Gyu; Kim, Sung Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    Nowadays, in Korea, advanced cladding such as FC92 is developed and its transient behaviors are required for the safety analysis of SFR. Design and safety analyses of sodium-cooled fast reactor (SFR) require understanding fuel pin responses to a wide range of off-normal events. In a loss-of-flow (LOF) or transient over-power (TOP), the temperature of the cladding is rapidly increased above its steady-state service temperature. Transient tests have been performed in sections of fuel pin cladding and a large data base has been established for austenitic stainless steel such as 20% cold-worked 316 SS and ferritic/martensitic steels such as HT9. This paper summarizes the technical status of transient testing facilities and their results. Previous researches showed the transient behaviors of HT9 cladding. For the safety analyses in SFR in Korea, simulated transient tests with newly developed FC92 as well as HT9 cladding are being carried out.

  5. First analysis of AGS0, LT2 and E9 CABRI tests with the new SFR safety code ASTEC-Na

    International Nuclear Information System (INIS)

    Perez-Martin, Sara; Bandini, Giacomino; Matuzas, Vaidas; Buck, Michael; Girault, Nathalie

    2015-01-01

    Within the framework of the European JASMIN project, the ASTEC-Na code is being developed for safety analysis of severe accidents in SFR. In the first phase of validation of the ASTEC-Na fuel thermo-mechanical models three in-pile tests conducted in the CABRI experimental reactor have been selected to be analysed. We present here the preliminary results of the simulation of two Transient Over Power tests and one power ramp test (AGS0, LT2 and E9, respectively) where no pin failure occurred during the transient. We present the comparison of ASTEC-Na results against experimental data and other safety code results for the initial steady state conditions prior to the transient onset as well as for the fuel pin behaviour during the transients. (author)

  6. Benchmarking LWR codes capability to model radionuclide deposition within SFR containments: An analysis of the Na ABCOVE tests

    International Nuclear Information System (INIS)

    Herranz, Luis E.; Garcia, Monica; Morandi, Sonia

    2013-01-01

    Highlights: • Assessment of LWR codes capability to model aerosol deposition within SFR containments. • Original hypotheses proposed to partially accommodate drawbacks from Na oxidation reactions. • A defined methodology to derive a more accurate characterization of Na-based particles. • Key missing models in LWR codes for SFR applications are identified. - Abstract: Postulated BDBAs in SFRs might result in contaminated-coolant discharge at high temperature into the containment. A full scope safety analysis of this reactor type requires computation tools properly validated in all the related fields. Radionuclide transport, particularly within the containment, is one of those fields. This sets two major challenges: to have reliable codes available and to build up a sound data base. Development of SFR source term codes was abandoned in the 80's and few data are available at present. The ABCOVE experimental programme conducted in the 80's is still a reference in the field. Postulated BDBAs in SFRs might result in contaminated-coolant discharge at high temperature into the containment. A full scope safety analysis of this reactor type requires computation tools properly validated in all the related fields. Radionuclide deposition, particularly within the containment, is one of those fields. This sets two major challenges: to have reliable codes available and to build up a sound data base. Development of SFR source term codes was abandoned in the 80's and few data are available at present. The ABCOVE experimental programme conducted in the 80's is still a reference in the field. The present paper is aimed at assessing the current capability of LWR codes to model aerosol deposition within a SFR containment under BDBA conditions. Through a systematic application of the ASTEC, ECART and MELCOR codes to relevant ABCOVE tests, insights have been gained into drawbacks and capabilities of these computation tools. Hypotheses and approximations have been adopted so that

  7. Project SAFE. Update of the SFR-1 safety assessment. Phase 1. Appendix A5: Radionuclide transport

    International Nuclear Information System (INIS)

    Moreno, L.

    1998-01-01

    A critical revision of the previous safety assessments made by SKB on the Final Repository for Radioactive Operational Waste, SFR is presented. The review of the Deepened Safety Assessment is also discussed. Based on this critical revision improvements are suggested. Hydrology, formation of complexes, and long-term behaviour of the barriers are some of the aspects where the safety assessment could be improved

  8. Discussion on safety analysis approach for sodium fast reactors

    International Nuclear Information System (INIS)

    Hong, Soon Joon; Choo, Yeon Joon; Suh, Nam Duk; Shin, Ahn Dong; Bae, Moo Hoon

    2012-01-01

    Utilization of nuclear energy is increasingly necessary not only because of the increasing energy consumption but also because of the controls on greenhouse emissions against global warming. To keep step with such demands, advanced reactors are now world widely under development with the aims of highly economical advances, and enhanced safety. Recently, further elaborating is encouraged on the research and development program for Generation IV (GEN IV) reactors, and in collaboration with other interested countries through the Generation IV International Forum (GIF). Sodium cooled Fast Reactor (SFR) is a strong contender amongst the GEN IV reactor concepts. Korea also takes part in that program and plans to construct demonstration reactor of SFR. SFR is under the development for a candidate of small modular reactors, for example, PRISM (Power Reactor Innovative Small Module). Understanding of safety analysis approach has also advanced by the demand of increasing comprehensive safety requirement. Reviewing the past development of the licensing and safety basis in the advanced reactors, such approaches seemed primarily not so satisfactory because the reference framework of licensing and safety analysis approach in the advanced reactors was always the one in water reactors. And, the framework is very plant specific one and thereby the advanced reactors and their frameworks don't look like a well assorted couple. Recently as a result of considerable advances in probabilistic safety assessment (PSA), risk informed approaches are increasingly applied together with some of the deterministic approaches like as the ones in water reactors. Technology neutral framework (TNF) can be said to be the utmost works of such risk informed approaches, even though an intensive assessment of the applicability has not been sufficiently accomplished. This study discusses the viable safety analysis approaches for the urgent application to the construction of pool type SFR. As discussed in

  9. Benchmarking LWR codes capability to model radionuclide deposition within SFR containments: An analysis of the Na ABCOVE tests

    Energy Technology Data Exchange (ETDEWEB)

    Herranz, Luis E., E-mail: luisen.herranz@ciemat.es [CIEMAT, Unit of Nuclear Safety Research, Av. Complutense, 40, 28040 Madrid (Spain); Garcia, Monica, E-mail: monica.gmartin@ciemat.es [CIEMAT, Unit of Nuclear Safety Research, Av. Complutense, 40, 28040 Madrid (Spain); Morandi, Sonia, E-mail: sonia.morandi@rse-web.it [Nuclear and Industrial Plant Safety Team, Power Generation System Department, RSE, via Rubattino 54, 20134 Milano (Italy)

    2013-12-15

    Highlights: • Assessment of LWR codes capability to model aerosol deposition within SFR containments. • Original hypotheses proposed to partially accommodate drawbacks from Na oxidation reactions. • A defined methodology to derive a more accurate characterization of Na-based particles. • Key missing models in LWR codes for SFR applications are identified. - Abstract: Postulated BDBAs in SFRs might result in contaminated-coolant discharge at high temperature into the containment. A full scope safety analysis of this reactor type requires computation tools properly validated in all the related fields. Radionuclide transport, particularly within the containment, is one of those fields. This sets two major challenges: to have reliable codes available and to build up a sound data base. Development of SFR source term codes was abandoned in the 80's and few data are available at present. The ABCOVE experimental programme conducted in the 80's is still a reference in the field. Postulated BDBAs in SFRs might result in contaminated-coolant discharge at high temperature into the containment. A full scope safety analysis of this reactor type requires computation tools properly validated in all the related fields. Radionuclide deposition, particularly within the containment, is one of those fields. This sets two major challenges: to have reliable codes available and to build up a sound data base. Development of SFR source term codes was abandoned in the 80's and few data are available at present. The ABCOVE experimental programme conducted in the 80's is still a reference in the field. The present paper is aimed at assessing the current capability of LWR codes to model aerosol deposition within a SFR containment under BDBA conditions. Through a systematic application of the ASTEC, ECART and MELCOR codes to relevant ABCOVE tests, insights have been gained into drawbacks and capabilities of these computation tools. Hypotheses and approximations have

  10. X447 EBR-II Experiment Benchmark for Verification of Audit Code of SFR Metal Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Yong Won; Bae, Moo-Hoon; Shin, Andong; Suh, Namduk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2016-10-15

    In KINS (Korea Institute of Nuclear Safety), to prepare audit calculation of PGSFR licensing review, the project has been started to develop the regulatory technology for SFR system including a fuel area. To evaluate the fuel integrity and safety during an irradiation, the fuel performance code must be used for audit calculation. In this study, to verify the new code system, the benchmark analysis is performed. In the benchmark, X447 EBR-II experiment data are used. Additionally, the sensitivity analysis according to mass flux change of coolant is performed. In case of LWR fuel performance modeling, various and advanced models have been proposed and validated based on sufficient in-reactor test results. However, due to the lack of experience of SFR operation, the current understanding of SFR fuel behavior is limited. In this study, X447 EBR-II Experiment data are used for benchmark. The fuel composition of X447 assembly is U-10Zr and PGSFR also uses this composition in initial phase. So we select X447 EBR-II experiment for benchmark analysis. Due to the lack of experience of SFR operation and data, the current understanding of SFR fuel behavior is limited. However, in order to prepare the licensing of PGSFR, regulatory audit technologies of SFR must be secured. So, in this study, to verify the new audit fuel performance analysis code, the benchmark analysis is performed using X447 EBR-II experiment data. Also, the sensitivity analysis with mass flux change of coolant is performed. In terms of verification, it is considered that the results of benchmark and sensitivity analysis are reasonable.

  11. X447 EBR-II Experiment Benchmark for Verification of Audit Code of SFR Metal Fuel

    International Nuclear Information System (INIS)

    Choi, Yong Won; Bae, Moo-Hoon; Shin, Andong; Suh, Namduk

    2016-01-01

    In KINS (Korea Institute of Nuclear Safety), to prepare audit calculation of PGSFR licensing review, the project has been started to develop the regulatory technology for SFR system including a fuel area. To evaluate the fuel integrity and safety during an irradiation, the fuel performance code must be used for audit calculation. In this study, to verify the new code system, the benchmark analysis is performed. In the benchmark, X447 EBR-II experiment data are used. Additionally, the sensitivity analysis according to mass flux change of coolant is performed. In case of LWR fuel performance modeling, various and advanced models have been proposed and validated based on sufficient in-reactor test results. However, due to the lack of experience of SFR operation, the current understanding of SFR fuel behavior is limited. In this study, X447 EBR-II Experiment data are used for benchmark. The fuel composition of X447 assembly is U-10Zr and PGSFR also uses this composition in initial phase. So we select X447 EBR-II experiment for benchmark analysis. Due to the lack of experience of SFR operation and data, the current understanding of SFR fuel behavior is limited. However, in order to prepare the licensing of PGSFR, regulatory audit technologies of SFR must be secured. So, in this study, to verify the new audit fuel performance analysis code, the benchmark analysis is performed using X447 EBR-II experiment data. Also, the sensitivity analysis with mass flux change of coolant is performed. In terms of verification, it is considered that the results of benchmark and sensitivity analysis are reasonable

  12. 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

  13. Germany: Assessment of the efficiency of a passive safety system for prevention of severe accidents for SFR

    International Nuclear Information System (INIS)

    Bubelis, E.

    2015-01-01

    The aim of the study was the evaluation of severe transient behavior in Sodium-cooled Fast Reactor (SFR) and of the impact of newly conceived inherent mitigation measures (the use of ASD – additional shutdown device). The SFR design taken for the analysis was the SFR(v2b-ST) reactor design, and the system code to be used was selected to be the SIM-SFR code. The transients chosen for evaluation of the efficiency of mitigation measures were the unprotected loss-of-flow (ULOF) and the unprotected loss-of-heat-sink (ULOHS)

  14. State of the art of CATHARE model for transient safety analysis of ASTRID SFR

    International Nuclear Information System (INIS)

    Lavastre, R.; Conti, A.; Marsault, Ph.; Chenaud, M.S.; Tosello, A.

    2014-01-01

    Within the framework of the ASTRID project (Advanced Sodium Technological Reactor for Industrial Demonstration), the conceptual design studies are being conducted in accordance with the GEN IV reactor objectives, particularly in terms of improving safety. This involves enhancing the general design in order to : - increase the safety margins for all unprotected-loss-of-flow (ULOF) and unprotected-loss-of-heat-sink (ULOHS) transients, - identify the need for additional safety devices that would complement core natural behavior so that temperature criteria on coolant, core and primary circuit structures can remain under the safety criteria. For this purpose, the use of CATHARE system code has been very important from the early stage of design in order to ensure a feedback for design teams to improve behavior during unprotected transients. Until 2012, CATHARE ULOxx transient calculations have been used mainly to compare different core designs. They contributed to lead to the choice of CFV core (axially heterogeneous core with an upper sodium plenum employed to achieve a negative sodium void reactivity worth). Meanwhile, models for an accurate core description and transients have been developed in CATHARE to improve the calculations towards best estimate calculations for safety analysis. This paper therefore presents these main developments in core modeling achieved for the 2 past years. For instance, we will focus on the way of dealing with fuel assemblies that have to be grouped together in the CATHARE code to form a channel with similar neutronic physics and thermal-hydraulics characteristics. We will also explain the way we deal with heterogeneity of fuel pin to obtain the accurate fuel temperature along the axis and to take into account pellet-cladding gap state. These two points have a great importance on feedback effects linked to the fuel, mainly the Doppler effect. The paper will finally introduce the upcoming improvements that are under development nowadays

  15. Model summary report for the safety assessment SFR 1 SAR-08

    Energy Technology Data Exchange (ETDEWEB)

    2008-03-15

    This document is the model summary report for the safety assessment SFR 1 SAR-08. In the report, the quality assurance measures conducted for the assessment codes are presented together with the chosen methodology. In the safety assessment SFR1 SAR-08, a number of different computer codes are used. In order to better understand how these codes are related an Assessment Model Flowchart, AMF, has been produced within the project. From the AMF, it is possible to identify the different modelling tasks and consequently also the different computer codes used. A number of different computer codes are used in the assessment of which some are commercial while others are developed for assessment projects. QA requirements must on the one hand take this diversity into account and on the other hand be well defined. In the methodology section of the report the following requirements are defined: - It must be demonstrated that the code is suitable for its purpose. - It must be demonstrated that the code has been properly used. - It must be demonstrated that the code development process has followed appropriate procedures and that the code produces accurate results. Although the requirements are identical for all codes, the measures used to show that the requirements are fulfilled will be different for different codes (for instance due to the fact that for some software the source-code is not available for review). Subsequent to the methodology section, each assessment code is presented and it is shown how the requirements are met

  16. Model summary report for the safety assessment SFR 1 SAR-08

    International Nuclear Information System (INIS)

    2008-03-01

    This document is the model summary report for the safety assessment SFR 1 SAR-08. In the report, the quality assurance measures conducted for the assessment codes are presented together with the chosen methodology. In the safety assessment SFR1 SAR-08, a number of different computer codes are used. In order to better understand how these codes are related an Assessment Model Flowchart, AMF, has been produced within the project. From the AMF, it is possible to identify the different modelling tasks and consequently also the different computer codes used. A number of different computer codes are used in the assessment of which some are commercial while others are developed for assessment projects. QA requirements must on the one hand take this diversity into account and on the other hand be well defined. In the methodology section of the report the following requirements are defined: - It must be demonstrated that the code is suitable for its purpose. - It must be demonstrated that the code has been properly used. - It must be demonstrated that the code development process has followed appropriate procedures and that the code produces accurate results. Although the requirements are identical for all codes, the measures used to show that the requirements are fulfilled will be different for different codes (for instance due to the fact that for some software the source-code is not available for review). Subsequent to the methodology section, each assessment code is presented and it is shown how the requirements are met

  17. Metallic Reactor Fuel Fabrication for SFR

    Energy Technology Data Exchange (ETDEWEB)

    Song, Hoon; Kim, Jong-Hwan; Ko, Young-Mo; Woo, Yoon-Myung; Kim, Ki-Hwan; Lee, Chan-Bock [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    The metal fuel for an SFR has such advantages such as simple fabrication procedures, good neutron economy, high thermal conductivity, excellent compatibility with a Na coolant, and inherent passive safety 1. U-Zr metal fuel for SFR is now being developed by KAERI as a national R and D program of Korea. The fabrication technology of metal fuel for SFR has been under development in Korea as a national nuclear R and D program since 2007. The fabrication process for SFR fuel is composed of (1) fuel slug casting, (2) loading and fabrication of the fuel rods, and (3) fabrication of the final fuel assemblies. Fuel slug casting is the dominant source of fuel losses and recycled streams in this fabrication process. Fabrication on the rod type metallic fuel was carried out for the purpose of establishing a practical fabrication method. Rod-type fuel slugs were fabricated by injection casting. Metallic fuel slugs fabricated showed a general appearance was smooth.

  18. Project SFR 1 SAR-08. Update of priority of FEPs from Project SAFE

    International Nuclear Information System (INIS)

    Gordon, Anna; Loefgren, Martin; Lindgren, Maria

    2008-03-01

    SFR 1 is a repository for final disposal of low and intermediate level radioactive waste produced at Swedish nuclear power plants, as well as at Swedish industrial, research, and medical treatment facilities. The repository obtained operational license in March 1988. The aim of Project SFR 1 SAR-08 is to perform an updated safety analysis, according to requirements in the regulations. A major difference between this and previous safety analyses is that repository safety should be demonstrated for 100,000 years after repository closure. This should be compared with the time frame of the safety assessment in Project SAFE that was 10,000 years. Due to the extended time frame, permafrost and glaciation have to be considered in the reference evolution of Project SFR 1 SAR-08. Other rationales for the update are recent input from the authorities concerning SAFE documents and the SFR 1 repository, as well as new data concerning the SFR 1 inventory. This report describes the outcome of revisiting the qualitative FEP (Features, Events and Processes) analysis carried out within Project SAFE for the SFR 1 repository. Each and every interaction definition, as defined in SAFE, has been examined with the aim at assuring that the SAFE interaction matrices are also applicable for SAR-08. It was found that this is generally the case, but seven new interactions were defined in order to make the interaction matrices more applicable for SAR-08. The priority of all interactions assigned priority 1 and many interactions assigned priority 2 in SAFE have been carefully examined. The examination has been made in the context of the general initial and boundary conditions that should also form the basis for the SAR-08 main scenario and less probable scenarios. In 48 cases, the priority of the interaction needed upgrading, compared to in SAFE. In a majority of these cases, the upgrade is due to the extended time frame of the safety assessment, from 10,000 years in SAFE to 100,000 years in SAR

  19. Project SFR 1 SAR-08. Update of priority of FEPs from Project SAFE

    Energy Technology Data Exchange (ETDEWEB)

    Gordon, Anna (Swedish Nuclear Fuel and Waste Management Co., Stockholm (SE)); Loefgren, Martin; Lindgren, Maria (Kemakta Konsult AB, Stockholm (SE)) (eds.)

    2008-03-15

    SFR 1 is a repository for final disposal of low and intermediate level radioactive waste produced at Swedish nuclear power plants, as well as at Swedish industrial, research, and medical treatment facilities. The repository obtained operational license in March 1988. The aim of Project SFR 1 SAR-08 is to perform an updated safety analysis, according to requirements in the regulations. A major difference between this and previous safety analyses is that repository safety should be demonstrated for 100,000 years after repository closure. This should be compared with the time frame of the safety assessment in Project SAFE that was 10,000 years. Due to the extended time frame, permafrost and glaciation have to be considered in the reference evolution of Project SFR 1 SAR-08. Other rationales for the update are recent input from the authorities concerning SAFE documents and the SFR 1 repository, as well as new data concerning the SFR 1 inventory. This report describes the outcome of revisiting the qualitative FEP (Features, Events and Processes) analysis carried out within Project SAFE for the SFR 1 repository. Each and every interaction definition, as defined in SAFE, has been examined with the aim at assuring that the SAFE interaction matrices are also applicable for SAR-08. It was found that this is generally the case, but seven new interactions were defined in order to make the interaction matrices more applicable for SAR-08. The priority of all interactions assigned priority 1 and many interactions assigned priority 2 in SAFE have been carefully examined. The examination has been made in the context of the general initial and boundary conditions that should also form the basis for the SAR-08 main scenario and less probable scenarios. In 48 cases, the priority of the interaction needed upgrading, compared to in SAFE. In a majority of these cases, the upgrade is due to the extended time frame of the safety assessment, from 10,000 years in SAFE to 100,000 years in SAR

  20. Project SFR 1 SAR-08. Update of priority of FEPs from Project SAFE

    Energy Technology Data Exchange (ETDEWEB)

    Gordon, Anna [Swedish Nuclear Fuel and Waste Management Co., Stockholm (SE); Loefgren, Martin; Lindgren, Maria [Kemakta Konsult AB, Stockholm (SE); eds.

    2008-03-15

    SFR 1 is a repository for final disposal of low and intermediate level radioactive waste produced at Swedish nuclear power plants, as well as at Swedish industrial, research, and medical treatment facilities. The repository obtained operational license in March 1988. The aim of Project SFR 1 SAR-08 is to perform an updated safety analysis, according to requirements in the regulations. A major difference between this and previous safety analyses is that repository safety should be demonstrated for 100,000 years after repository closure. This should be compared with the time frame of the safety assessment in Project SAFE that was 10,000 years. Due to the extended time frame, permafrost and glaciation have to be considered in the reference evolution of Project SFR 1 SAR-08. Other rationales for the update are recent input from the authorities concerning SAFE documents and the SFR 1 repository, as well as new data concerning the SFR 1 inventory. This report describes the outcome of revisiting the qualitative FEP (Features, Events and Processes) analysis carried out within Project SAFE for the SFR 1 repository. Each and every interaction definition, as defined in SAFE, has been examined with the aim at assuring that the SAFE interaction matrices are also applicable for SAR-08. It was found that this is generally the case, but seven new interactions were defined in order to make the interaction matrices more applicable for SAR-08. The priority of all interactions assigned priority 1 and many interactions assigned priority 2 in SAFE have been carefully examined. The examination has been made in the context of the general initial and boundary conditions that should also form the basis for the SAR-08 main scenario and less probable scenarios. In 48 cases, the priority of the interaction needed upgrading, compared to in SAFE. In a majority of these cases, the upgrade is due to the extended time frame of the safety assessment, from 10,000 years in SAFE to 100,000 years in SAR

  1. Project SAFE. Update of the SFR-1 safety assessment. Phase 1. Appendix A1: Inventory

    International Nuclear Information System (INIS)

    Riggare, P.

    1998-01-01

    One of the aims in the safety assessment of SFR-1 is to estimate the release to the environment. In order to make these calculations there is a need to describe the inventory in greater detail. The new computerised database of waste in SFR-1 gives a good possibility to achieve this. The aim for project SAFE is to make both conservative and realistic radionuclide transport calculations. To achieve this goal there must be two inventories. The conservative inventory is the inventory used in the design of the repository, which in most parts is identical with the limits in the licence for SFR-1. There is a great interest to have good estimates of the volumes of the different waste types. A thorough prognosis should be made in 1999, but until then the latest one from 1995 could be used in the calculations. The total (actual) inventory of nuclides is calculated from the measurements of the easy-to-measure nuclides since, in principle, all hard-to-measure nuclides are calculated by correlation factors to 60 Co and 137 Cs . These factors should be reviewed since there are quite large uncertainties involved. 14 C dominates the individual doses after a few hundred years and the collective dose in the inland-scenario. The amount of the nuclide is uncertain since the correlation factor is very uncertain. The chemical speciation of 14 C is also of interest due to different properties of organic and inorganic carbon. 36 Cl is very hard to measure. Although the authorities in their reviews of the safety reports say that there probably are small doses from chlorine, the inventory should be improved. 59 Ni is a long-lived nuclide that sets a limit to the close-to-the-core metal scrap that can be taken to SFR- 1. There is an ongoing research project to provide a better measuring method for 59 Ni. This should make it possible to improve the knowledge about 59 Ni inventory. The assumption that 90 % of the inventory is collected in the ion-exchange resins should be checked. Actinides

  2. Site investigation SFR. Hydrogeological modelling of SFR. Model version 0.2

    Energy Technology Data Exchange (ETDEWEB)

    Oehman, Johan (Golder Associates AB (Sweden)); Follin, Sven (SF GeoLogic (Sweden))

    2010-01-15

    The Swedish Nuclear Fuel and Waste Management Company (SKB) has conducted site investigations for a planned extension of the existing final repository for short-lived radioactive waste (SFR). A hydrogeological model is developed in three model versions, which will be used for safety assessment and design analyses. This report presents a data analysis of the currently available hydrogeological data from the ongoing Site Investigation SFR (KFR27, KFR101, KFR102A, KFR102B, KFR103, KFR104, and KFR105). The purpose of this work is to develop a preliminary hydrogeological Discrete Fracture Network model (hydro-DFN) parameterisation that can be applied in regional-scale modelling. During this work, the Geologic model had not yet been updated for the new data set. Therefore, all analyses were made to the rock mass outside Possible Deformation Zones, according to Single Hole Interpretation. Owing to this circumstance, it was decided not to perform a complete hydro-DFN calibration at this stage. Instead focus was re-directed to preparatory test cases and conceptual questions with the aim to provide a sound strategy for developing the hydrogeological model SFR v. 1.0. The presented preliminary hydro-DFN consists of five fracture sets and three depth domains. A statistical/geometrical approach (connectivity analysis /Follin et al. 2005/) was performed to estimate the size (i.e. fracture radius) distribution of fractures that are interpreted as Open in geologic mapping of core data. Transmissivity relations were established based on an assumption of a correlation between the size and evaluated specific capacity of geologic features coupled to inflows measured by the Posiva Flow Log device (PFL-f data). The preliminary hydro-DFN was applied in flow simulations in order to test its performance and to explore the role of PFL-f data. Several insights were gained and a few model technical issues were raised. These are summarised in Table 5-1

  3. Site investigation SFR. Hydrogeological modelling of SFR. Model version 0.2

    International Nuclear Information System (INIS)

    Oehman, Johan; Follin, Sven

    2010-01-01

    The Swedish Nuclear Fuel and Waste Management Company (SKB) has conducted site investigations for a planned extension of the existing final repository for short-lived radioactive waste (SFR). A hydrogeological model is developed in three model versions, which will be used for safety assessment and design analyses. This report presents a data analysis of the currently available hydrogeological data from the ongoing Site Investigation SFR (KFR27, KFR101, KFR102A, KFR102B, KFR103, KFR104, and KFR105). The purpose of this work is to develop a preliminary hydrogeological Discrete Fracture Network model (hydro-DFN) parameterisation that can be applied in regional-scale modelling. During this work, the Geologic model had not yet been updated for the new data set. Therefore, all analyses were made to the rock mass outside Possible Deformation Zones, according to Single Hole Interpretation. Owing to this circumstance, it was decided not to perform a complete hydro-DFN calibration at this stage. Instead focus was re-directed to preparatory test cases and conceptual questions with the aim to provide a sound strategy for developing the hydrogeological model SFR v. 1.0. The presented preliminary hydro-DFN consists of five fracture sets and three depth domains. A statistical/geometrical approach (connectivity analysis /Follin et al. 2005/) was performed to estimate the size (i.e. fracture radius) distribution of fractures that are interpreted as Open in geologic mapping of core data. Transmissivity relations were established based on an assumption of a correlation between the size and evaluated specific capacity of geologic features coupled to inflows measured by the Posiva Flow Log device (PFL-f data). The preliminary hydro-DFN was applied in flow simulations in order to test its performance and to explore the role of PFL-f data. Several insights were gained and a few model technical issues were raised. These are summarised in Table 5-1

  4. Exploration of Important Issues for the Safety of SFR 1 using Performance Assessment Calculations

    International Nuclear Information System (INIS)

    Maul, P.R.; Robinson, P.C.

    2002-06-01

    SKB has produced a revised safety case for the SFR 1 disposal facility for low and intermediate level radioactive wastes at Forsmark: project SAFE. This assessment includes a Performance Assessment (PA) for the long term post-closure safety of the facility. SKI has a responsibility to scrutinise SKB's safety case that is shared with SSI. Quintessa has undertaken a review of SKB's case for the long term safety of SFR 1 to assist SKI's evaluation of SAFE, and this is given in SKI-R--02-61, henceforth referred to as the Quintessa Review. The current report describes the independent PA calculations that provided an input to that review. Since 1999 SKI has been developing a PA capability for SFR 1 using the AMBER software. Two key features of the approach taken have been: To represent the whole system in a single model; and To allow the time-dependency of all key features, events and processes to be represented. These capabilities allow a better understanding of the key features of the system to be obtained for different future evolutions (scenarios). This report presents a summary of the work undertaken to provide SKI with a PA capability for SFR 1 and the calculations undertaken with it. Calculations have been undertaken for radionuclides transported in groundwater and gas, but not for direct intrusion by humans into the wastes. It should be emphasised that the purpose of the Performance Assessment calculations described in this report is not to provide an alternative assessment of potential radiological impacts to that produced by SKB. The aim is to use the models that have been developed to investigate the important features of the system and to help SKI scrutinise the case put to them by SKB. The PA calculations that have been undertaken are by no means comprehensive, and various issues could be investigated further if required. The key issues that have been identified can be summarised as follows: 1. The SFR 1 system has a number of different timescales that can

  5. Structure of fuel performance audit code for SFR metal fuel

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Yong Sik; Kim, Hyo Chan [KAERI, Daejeon (Korea, Republic of); Jeong, Hye Dong; Shin, An Dong; Suh, Nam Duk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2012-10-15

    A Sodium Cooled Fast Reactor (SFR) is a promising option to solve the spent fuel problems, but, there are still much technical issues to commercialize a SFR. One of issues is a development of advanced fuel which can solve the safety and the economic issues at the same time. Since a nuclear fuel is the first barrier to protect radioactive isotope release, the fuel's integrity must be secured. In Korea Institute of Nuclear Safety (KINS), the new project has been started to develop the regulatory technology for SFR system including a fuel area. To evaluate the fuel integrity and safety during an irradiation, the fuel performance code must be used for audit calculation. To develop the new code system, the code structure design and its requirements need to be studied. Various performance models and code systems are reviewed and their characteristics are analyzed in this paper. Based on this study, the fundamental performance models are deduced and basic code requirements and structure are established.

  6. Project SAFE. Update of the SFR-1 safety assessment. Phase 1. Appendix A4: Far-field

    International Nuclear Information System (INIS)

    Follin, S.; Andersson, Johan; Holmen, J.; Axelsson, C.L.

    1998-01-01

    This appendix has identified potential needs for updated hydrogeological modelling of the SFR in connection to the planned update of the performance assessment of the SFR within the framework of the SAFE-project. The objectives of such updated modelling should be to present a credible representation of the hydrogeological system, to explore effects of seals and repository extensions and to provide input to the release and transport calculations of the assessment. The last objective has led to the conclusion that an important focus of the modelling should be to determine the flow through the vaults under different conditions as this flow appear to be a very important quantity in the radionuclide release calculations. The suggested modelling should use relevant data and apply modern modelling tools and techniques, but should be geared towards the objectives. For this reasons it is suggested to apply a set of complementary and sometimes nested approaches, where each model approach is set up in order to address a specific set of questions. Answering these questions would motivate simplifications made in subsequent steps of the modelling. To the extent possible the models should be compared with existing data on flow and Baltic water breakthrough. However, in making such comparisons the accuracy of the measurements and the precision of the models need to be considered. A one-to-one match cannot be expected. It appears that careful geochemical evaluation of the site would only be necessary if more credit is placed on migration in the geosphere. If such an evaluation is considered it should be co-ordinated with the regional groundwater modelling. The issue of gas production should be reconsidered in a scenario and process analysis of SFR. However, given the strong conclusions already made it appears that gas migration in the rock will still remain as a minor issue. The major assumptions going into the analysis of the near-field in the final safety report and the deepened

  7. Assessment of SFR reactor safety issues: Part II: Analysis results of ULOF transients imposed on a variety of different innovative core designs with SAS-SFR

    Energy Technology Data Exchange (ETDEWEB)

    Kruessmann, R., E-mail: regina.kruessmann@kit.edu [Karlsruhe Institute of Technology, Institute of Neutron Physics and Reactor Technology INR, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Ponomarev, A.; Pfrang, W.; Struwe, D. [Karlsruhe Institute of Technology, Institute of Neutron Physics and Reactor Technology INR, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Champigny, J.; Carluec, B. [AREVA, 10, rue J. Récamier, 69456 Lyon Cedex 06 (France); Schmitt, D.; Verwaerde, D. [EDF R& D, 1 avenue du général de Gaulle, 92140 Clamart (France)

    2015-04-15

    Highlights: • Comparison of different core designs for a sodium-cooled fast reactor. • Safety assessment with the code system SAS-SFR. • Unprotected Loss of Flow (ULOF) scenario. • Sodium boiling and core melting cannot be avoided. • A net negative Na void effect provides more grace time prior to local SA destruction. - Abstract: In the framework of cooperation agreements between KIT-INR and AREVA SAS NP as well as between KIT-INR and EDF R&D in the years 2008–2013, the evaluation of severe transient behavior in sodium-cooled fast reactors (SFRs) was investigated. In Part I of this contribution, the efficiency of newly conceived prevention and mitigation measures was investigated for unprotected loss-of-flow (ULOF), unprotected loss-of-heat-sink (ULOHS) and the unprotected transient-overpower (UTOP) transients. In this second part, consequence analyses were performed for the initiation phase of different unprotected loss-of-flow (ULOF) scenarios imposed on a variety of different core design options of SFRs. The code system SAS-SFR was used for this purpose. Results of analyses for cases postulating unavailability of prevention measures as shut-down systems, passive and/or active additional devices show that entering into an energetic power excursion as a consequence of the initiation phase of a ULOF cannot be avoided for those core designs with a cumulative void reactivity feedback larger than zero. However, even for core designs aiming at values of the void reactivity less than zero it is difficult to find system design characteristics which prevent the transient entering into partial core destruction. Further studies of the transient core and system behavior would require codes dedicated to specific aspects of transition phase analyses and of in-vessel material relocation analyses.

  8. Heat transfer analysis to investigate the core catcher plate assembly in SFR

    International Nuclear Information System (INIS)

    Patil, Swapnil; Sharma, Anil Kumar; Velusamy, K.; Nashine, B.K.; Selvaraj, P.

    2015-01-01

    Severe accident scenario in Sodium Cooled Fast Reactor (SFR) is the major concern for public acceptance. After severe accident, the molten core continuously generates substantial decay heat. However, an in-vessel core catcher plate is provided to remove the decay heat passively. The numerical investigation of pool hydraulics phenomena in sodium pool of typical Indian SFR has been carried out. The debris may form a heap with different angle over the core catcher plate due to molten fuel density and interaction force. Therefore, the debris bed with different heap angle has been analyzed for steady and transient state conditions. The governing equation of fluid flow and heat transfer are solved by finite volume method based solver with the k-ε turbulent model. The time period Δ for which temperature is exceeding above safety limit with different debris heap angle have been established. (author)

  9. The biosphere today and tomorrow in the SFR area

    Energy Technology Data Exchange (ETDEWEB)

    Kautsky, Ulrik (ed.)

    2001-06-01

    This report is a compilation of the work done mainly in the SAFE project for the biosphere from about 14 reports. The SAFE project is the updated safety analysis of SFR-1, the LLW and ILW repository at Forsmark. The aim of the report is to summarize the available information about the present-day biosphere in the area surrounding SFR and to use this information, together with information about the previous development of the biosphere, to predict the future development of the area in a more comparable way than the underlying reports. The data actually used for the models have been taken from the original reports which also justify or validate the data. The report compiles information about climate, oceanography, landscape, sedimentation, shoreline displacement, marine, lake and terrestrial ecosystems.

  10. The biosphere today and tomorrow in the SFR area

    International Nuclear Information System (INIS)

    Kautsky, Ulrik

    2001-06-01

    This report is a compilation of the work done mainly in the SAFE project for the biosphere from about 14 reports. The SAFE project is the updated safety analysis of SFR-1, the LLW and ILW repository at Forsmark. The aim of the report is to summarize the available information about the present-day biosphere in the area surrounding SFR and to use this information, together with information about the previous development of the biosphere, to predict the future development of the area in a more comparable way than the underlying reports. The data actually used for the models have been taken from the original reports which also justify or validate the data. The report compiles information about climate, oceanography, landscape, sedimentation, shoreline displacement, marine, lake and terrestrial ecosystems

  11. Long-time stability following freezing and thawing of concrete and bentonite in deposition of low- and intermediate-level radioactive waste in SFR 1

    International Nuclear Information System (INIS)

    Emborg, Mats; Jonasson, Jan-Erik; Knutsson, Sven

    2007-10-01

    This document describes the effect of freezing on the concrete and bentonite barriers in SFR 1. The document constitutes one of the references describing the degradation of barriers in a long-time perspective and is used in the safety analysis SFR 1 SAR-08

  12. Bedrock Hydrogeology-Site investigation SFR

    International Nuclear Information System (INIS)

    Oehman, Johan; Bockgaard, Niclas; Follin, Sven

    2012-06-01

    The Swedish Nuclear Fuel and Waste Management Company (SKB) has conducted site investigations for a planned extension of the existing final repository for short-lived radioactive waste (SFR). This report presents an integrated analysis and interpretation of the historic data from the existing SFR (1980 - 1986), as well as, from the recent investigations for the planned extension of SFR (2008 - 2009). The primary objective is to establish a conceptual hydrogeological model of the bedrock for safety assessment and design analyses. Analyses and interpretations of all (old and new) hydraulic data are analysed with regard to the recently developed geological deformation zone model of the SFR model domain (Curtis et al. 2011). The methodology used by Curtis et al. (2011) has focussed on magnetic anomalies and deformation zone intercepts with ground surface greater than 300 m. In the hydrogeological modelling, however, it has been considered important to also explore the occurrence and characteristics of shallow horizontal to sub-horizontal structures (sheet joints) inside the SFR model domain. Such structures are of considerable importance for the hydrogeology in the uppermost c. 150 m of bedrock in SDM-Site Forsmark; hence the term Shallow Bedrock Aquifer was used to emphasise their hydraulic significance. In this study, the acronym SBA-structure is used for horizontal structures identified in the hydrogeological modelling. In addition to the predominantly steeply dipping geological deformation zones, eight so-called SBA-structures are modelled deterministically in the hydrogeological model. The SBA-structures are envisaged as hydraulically heterogeneous and composed of clusters of minor gently dipping to horizontal fractures rather than extensive single features. A type of structures that is partly included in the definition of the SBA-structures is the Unresolved Possible Deformations Zone (Unresolved PDZ) intercepts identified by Curtis et al. (2011). The Unresolved

  13. Bedrock Hydrogeology - Site investigation SFR

    Energy Technology Data Exchange (ETDEWEB)

    Oehman, Johan [Geosigma AB, Stockholm (Sweden); Bockgaard, Niclas [Golder Assoes AB, Stockholm (Sweden); Follin, Sven [SF GeoLogic AB, Taeby (Sweden)

    2012-06-15

    The Swedish Nuclear Fuel and Waste Management Company (SKB) has conducted site investigations for a planned extension of the existing final repository for short-lived radioactive waste (SFR). This report presents an integrated analysis and interpretation of the historic data from the existing SFR (1980 - 1986), as well as, from the recent investigations for the planned extension of SFR (2008 - 2009). The primary objective is to establish a conceptual hydrogeological model of the bedrock for safety assessment and design analyses. Analyses and interpretations of all (old and new) hydraulic data are analysed with regard to the recently developed geological deformation zone model of the SFR model domain (Curtis et al. 2011). The methodology used by Curtis et al. (2011) has focussed on magnetic anomalies and deformation zone intercepts with ground surface greater than 300 m. In the hydrogeological modelling, however, it has been considered important to also explore the occurrence and characteristics of shallow horizontal to sub-horizontal structures (sheet joints) inside the SFR model domain. Such structures are of considerable importance for the hydrogeology in the uppermost c. 150 m of bedrock in SDM-Site Forsmark; hence the term Shallow Bedrock Aquifer was used to emphasise their hydraulic significance. In this study, the acronym SBA-structure is used for horizontal structures identified in the hydrogeological modelling. In addition to the predominantly steeply dipping geological deformation zones, eight so-called SBA-structures are modelled deterministically in the hydrogeological model. The SBA-structures are envisaged as hydraulically heterogeneous and composed of clusters of minor gently dipping to horizontal fractures rather than extensive single features. A type of structures that is partly included in the definition of the SBA-structures is the Unresolved Possible Deformations Zone (Unresolved PDZ) intercepts identified by Curtis et al. (2011). The Unresolved

  14. Analysis of Accident Scenarios for the Development of Probabilistic Safety Assessment Model for the Metallic Fuel Sodium-Cooled Fast Reactor

    International Nuclear Information System (INIS)

    Kim, Tae Woon; Park, S. Y.; Yang, J. E.; Kwon, Y. M.; Jeong, H. Y.; Suk, S. D.; Lee, Y. B.

    2009-03-01

    The safety analysis reports which were reported during the development of sodium cooled fast reactors in the foreign countries are reviewed for the establishment of Probabilistic Safety Analysis models for the domestic SFR which are under development. There are lots of differences in the safety characteristics between the mixed oxide (MOX) fuel SFR and metallic fuel SFR. Metallic fuel SFR is under development in Korea while MOX fuel SFR is under development in France, Japan, India and China. Therefore the status on the development of fast reactors in the foreign countries are reviewed at first and then the safety characteristics between the MOX fuel SFR and the metallic fuel SFR are reviewed. The core damage can be defined as coolant voiding, fuel melting, cladding damage. The melting points of metallic fuel and the MOX fuel is about 1000 .deg. C and 2300 .deg. C, respectively. The high energy stored in the MOX fuel have higher potential to voiding of coolant compared to the possibility in the metallic fuel. The metallic fuel has also inherent reactivity feedback characteristic that the metallic fuel SFR can be shutdown safely in the events of transient overpower, loss of flow, and loss of heat sink without scram. The metallic fuel has, however, lower melting point due to the eutectic formation between the uranium in metallic fuel and the ferrite in metallic cladding. It is needed to identify the core damage accident scenarios to develop Level-1 PSA model. SSC-K computer code is used to identify the conditions in which the core damage can occur in the KALIMER-600 SFR. The accident cases which are analyzed are the triple failure accidents such as unprotected transient over power events, loss of flow events, and loss of heat sink events with impaired safety systems or functions. Through the analysis of the triple failure accidents for the KALIMER-600 SFR, it is found that the PSA model developed for the PRISM reactor design can be applied to KALIMER-600. However

  15. Status of SFR Metal Fuel Development

    International Nuclear Information System (INIS)

    Lee, Chan Bock; Lee, Byoung Oon; Kim, Ki Hwan; Kim, Sung Ho

    2013-01-01

    Conclusion: • Metal fuel recycling in SFR: - Enhanced utilization of uranium resource; - Efficient transmutation of minor actinides; - Inherent passive reactor safety; - Proliferation resistance with pyro-electrochemical fuel recycling. • Demonstration of technical feasibility of recycling TRU metal fuel by 2020: - Remote fuel fabrication; - Irradiation performance up to high burnup

  16. Examining memorandum: Ultimate store for nuclear reactor wastes - SFR-1

    International Nuclear Information System (INIS)

    Bergman, C.; Ericsson, G.; Godaas, T.; Haegg, C.; Johansson, G.

    1988-01-01

    The report constitutes the basis for the position of the National Institute of Radiation Protection as regards permission to operate SFR-1 at Forsmark. The memorandum describes: - existing conditions regarding commissioning SFR-1, - summarily the final safety report from the Swedish Fuel and Waste Management Co, - consultant contributions ordered in connection with the examination, - the judgement of the institute in all questions relevant to radiation protection conditions in SFR-1. The institute has made it's own estimates of the radiation doses the repository could be the source of. It is concluded that the radiation doses from the repository are acceptable and consequently operation permission is recommended. (O.S.)

  17. Mechanical modelling of transient- to- failure SFR fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Feria, F.; Herranz, L. E.

    2014-07-01

    The response of Sodium Fast Reactor (SFR) fuel rods to transient accident conditions is an important safety concern. During transients the cladding strain caused by the stress due to pellet cladding mechanical interaction (PCMI) can lead to failure. Due to the fact that SFR fuel rods are commonly clad with strengthened material made of stainless steel (SS), cladding is usually treated as an elastic-perfectly-plastic material. However, viscoplastic behaviour can contribute to mechanical strain at high temperature (> 1000 K). (Author)

  18. Data for calibration and validation of numerical models at SFR Nuclear Waste Repository

    International Nuclear Information System (INIS)

    Axelsson, Carl-Lennart

    1997-12-01

    The final repository for low and intermediate radioactive waste, SFR, is located below the Baltic, offshore of the nuclear power plant at Forsmark. The current operating permit for SKB stipulates that the safety assessment is updated at least every ten year. In response, SKB has started the SAFE project which aims at submitting a complete revised safety analysis before or during the year 2000. The current report is part of the far-field analyses in SAFE and provides information that can be used in a revised hydrogeological modelling of the facility. Information have been collected mainly during the construction phase of SFR, 1983 - 88, and the operation phase from 1988. The specific information that is available for the construction phase is: pressure responses in different bore holes when pumping in one bore hole, groundwater pressure in sections of bore holes, inflow to different parts of the SFR, and groundwater chemistry and isotope analyses in sections of bore holes. During the operation phase, the following information is available: ground-water pressure in sections of bore holes, inflow to different parts of the SFR facility, and groundwater chemistry and isotope analyses in sections of bore holes. The important issues in the groundwater modelling for the performance assessment study of SFR is the amount of water that flows through the storage caverns and the silo together with the possible retention and adsorption in the rock mass, i.e. the flow paths from the repository parts. Thus, the most important type of information is the inflow measurements made in different parts of SFR. The groundwater chemistry may be used to understand the flow pattern and mixing of water with various origin such as fresh groundwater, saline rock/fracture groundwater and Baltic Sea water, especially to predict breakthrough time for the Baltic Sea water at different bore hole sections in fracture zones. The report discusses especially the availability and evolution of inflow and

  19. Preliminary Analysis of the Bundle-Duct Interaction for the fuel of SFR

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Byoung Oon; Cheon, Jin Sik; Hahn, Do Hee; Lee, Chan Bock [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-10-15

    BDI (Bundle-Duct Interaction) occurs in the fuel of SFR (Sodium-cooled Fast Reactor) due to the radial expansion and bowing of a fuel pin bundle. Under the BDI condition, excess cladding strain and hot spots would occur. Therefore, BDI, which is the dominant deformation mechanisms in a fuel pin bundle, should be considered to evaluate the FBR fuel integrity. The analysis codes such as ETOILE and BMBOO, have been developed to evaluate the BDI behavior. The bundle duct interaction model is also being developed for SFR in Korea. This model is based on ANSYS. In this paper, the fuel pin configuration model for the BDI calculation was established. The preliminary analysis of the bundle-duct interaction was performed to evaluate the fuel design concept.

  20. The SSI and SKI review of the updated Final Safety Report for SFR 1 issued by SKB. Review report; SSI:s och SKI:s granskning av SKB:s uppdaterade Slutlig Saekerhetsrapport foer SFR 1. Granskningsrapport

    Energy Technology Data Exchange (ETDEWEB)

    Avila, Rodolfo; Jensen, Mikael; Larsson, Carl-Magnus; Lund, Ingemar; Loefgren, Tomas; Moberg, Leif; Norden, Maria; Wiebert, Anders [Swedish Radiation Protection Authority, Stockholm (Sweden); Berglund, Thomas; Dverstorp, Bjoern; Hedberg, Bengt; Kautsky, Fritz; Lilja, Christina; Simic, Eva; Stroemberg, Bo; Sundstroem, Benny; Toverud, Oeivind; Wingefors, Stig; Zika, Helmuth [Swedish Nuclear Power Inspectorate, Stockholm (Sweden)

    2003-11-01

    The repository for operational radioactive wastes in Sweden, SFR1, has been the object for a new safety assessment study by SKB (The Swedish Nuclear Fuel and Waste Management Co.). The findings of the review group will form the basis for decisions by the authorities on the provisions for the future operation of the repository.

  1. The nuclide inventory in SFR-1; Nuklidinventariet i SFR-1

    Energy Technology Data Exchange (ETDEWEB)

    Ingemansson, Tor [ALARA Engineering, Skultuna (Sweden)

    2001-10-01

    This report is an account for a project carried out on behalf of the Swedish Radiation Protection Authority (SSI): 'Nuclide inventory in SFR-1' (The Swedish underground disposal facility for low and intermediate level reactor waste). The project comprises the following five sub-projects: 1) Measuring methods for nuclides, difficult to measure, 2) The nuclide inventory in SFR-1, 3) Proposal for nuclide library for SFR-1 and ground disposal, 4) Nuclide library for exemption, and 5) Characterising of the nuclide inventory and documentation for SFL waste. In all five sub-projects long-lived activity, including Cl-36, has been considered.

  2. Modelling of future hydrogeological conditions at SFR

    International Nuclear Information System (INIS)

    Holmen, L.G.; Stigsson, M.

    2001-03-01

    The purpose is to estimate the future groundwater movements at the SFR repository and to produce input to the quantitative safety assessment of the SFR. The future flow pattern of the groundwater is of interest, since components of the waste emplaced in a closed and abandoned repository will dissolve in the groundwater and be transported by the groundwater to the ground surface. The study is based on a system analysis approach. Three-dimensional models were devised of the studied domain. The models include the repository tunnels and the surrounding rock mass with fracture zones. The formal models used for simulation of the groundwater flow are three-dimensional mathematical descriptions of the studied hydraulic system. The studied domain is represented on four scales - regional, local, semi local and detailed - forming four models with different resolutions: regional, local, semi local and detailed models. The local and detailed models include a detailed description of the tunnel system at SFR and of surrounding rock mass and fracture zones. In addition, the detailed model includes description of the different structures that take place inside the deposition tunnels. At the area studied, the shoreline will retreat due to the shore level displacement; this process is included in the models. The studied period starts at 2000 AD and continues until a steady state like situation is reached for the surroundings of the SFR at ca 6000 AD. The models predict that as long as the sea covers the ground above the SFR, the regional groundwater flow as well as the flow in the deposition tunnels are small. However, due to the shore level displacement the shoreline (the sea) will retreat. Because of the retreating shoreline, the general direction of the groundwater flow at SFR will change, from vertical upward to a more horizontal flow; the size of the groundwater flow will be increased as well. The present layout of the SFR includes five deposition tunnels: SILO, BMA, BLA, BTF1

  3. Heterogeneous recycling in SFR core periphery

    International Nuclear Information System (INIS)

    Varaine, Frederic; Buiron, Laurent; Boucher, Lionel; Chabert, Christine

    2008-01-01

    development, based on the one hand on the solutions offered by the existing fleet (reprocessing, fabrication and NPP) and on the other hand on the solutions offered by the future reactors of fourth generation. The scenario study considers the French nuclear park with a constant nuclear energy demand at 430 TWhe / year. The current nuclear park is replaced between 2020 and 2050 by a mixed nuclear park: 66 % of Generation III EPR reactors and 33% of Generation IV SFR. From 2080 to 2100, the EPR are replaced by SFR. The Plutonium is recycled in the fissile part of the SFR core. The separation of the minor actinides at the reprocessing step starts in 2038. The minor actinides are recycled in the radial blankets of the SFR from 2040 (10% content of MA). Those calculations are performed by the COSI code. The results indicate that the minor actinides inventory can be stabilized with the heterogeneous mode of transmutation using minor actinides in the radial blankets of the SFR. A minor actinides rate around 10% in the radial blankets is sufficient with the condition to involve 100 % of the SFR in the transmutation process. The minor actinides multi-recycling on a depleted uranium oxide matrix in radial blankets of SFR showed good results in terms of transmutation performances. This heterogeneous model allows a massive minor actinides loading while having almost no consequence on the core safety parameters and core fuel management. Two MA enrichment targets have been studied: an ambitious 40% case and a more realistic 10% case. The design of such assembly has to deal with criteria implying multi-physics analysis. The 10% MA content seems a good balance between transmutation performances and back/front end impact (neutrons source, decay heat,..) compared to the 40% content. Investigations, such as dedicated experimental material and fuel irradiation programs, are under process at CEA to set a global vision of an optimized system that can answer all these problems. (authors)

  4. Preliminary Analysis of the Fuel Bundle Stiffness by ANSYS for SFR

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Byoung Oon; Cheon, Jin Sik; Hahn, Do Hee; Lee, Chan Bock [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-05-15

    In SFR (Sodium-cooled Fast Reactor) the temperature of the fuel pin is higher than that of the hexagonal duct, so the thermal expansion rate of the fuel bundle is higher than that of the duct. The neutron fluence and the fuel pin pressure are also increased according to the burnup. So the radial expansion and bowing of a fuel pin bundle would occur, and then fuel bundle would interact with a duct. This phenomenon is called bundle-to-duct interaction (BDI). Under the BDI condition, excess cladding strain and hot spots would occur. Therefore BDI as well as the core mechanics should be considered to evaluate the FBR fuel integrity. The analysis codes such as ETOILE, SHADOW, and MARSE, have been developed to evaluate the BDI behavior. The ANSYS based model is also being developed to analysis the bundle duct interaction for SFR in Korea. In this paper, the fuel pin/bundle model for analyzing the bending deflection and oval deformation was described. The preliminary analysis of the fuel bundle stiffness was performed by the developed model.

  5. SFR site investigation. Bedrock Hydrogeochemistry

    International Nuclear Information System (INIS)

    Nilsson, Ann-Chatrin; Tullborg, Eva-Lena; Smellie, John; Gimeno, Maria J.; Gomez, Javier B.; Auque, Luis F.; Sandstroem, Bjoern; Pedersen, Karsten

    2011-11-01

    There are plans that the final repository for low and intermediate level radioactive waste, SFR, located about 150 km north of Stockholm, will be extended. Geoscientific studies to define and characterise a suitable bedrock volume for the extended repository have been carried out from 2007 to 2011, and have included the drilling and evaluation of seven new core drilled and four percussion boreholes. These new data, together with existing data extending back to 1985, have been interpreted and modelled in order to provide the necessary information for safety assessment and repository design. This report presents the final hydrogeochemical site description for the SFR site, and will constitute a background report for the integrated site description (the SFR Site Descriptive Model version 1.0) together with corresponding reports from the geological and hydrogeological disciplines. Most of the hydrogeochemical data from the field investigations consist of major ions and isotopes together with sporadic gas, microbe and measured redox data. Despite the close proximity of the Forsmark site, few data from this source are of relevance because of the shallow nature of the SFR site, the fact that SFR is located beneath the Baltic Sea and also the drawdown/upconing impacts of its construction on the hydrogeochemistry. This artificially imposed dynamic flow system is naturally more prevalent along major deformation fracture zones of higher transmissivity, whilst lower transmissive fractures together with the less transmissive bedrock masses between major deformation zones, still retain some evidence of the natural groundwater mixing patterns established prior to the SFR construction. The groundwaters in the SFR dataset cover a depth down to -250 m.a.s.l. with single sampling locations at -300 and -400 m.a.s.l. and represent a relatively limited salinity range (1,500 to 5,500 mg/L chloride). However, the δ 18 O values show a wide variation (-15.5 to -7.5 per mille V

  6. SFR site investigation. Bedrock Hydrogeochemistry

    Energy Technology Data Exchange (ETDEWEB)

    Nilsson, Ann-Chatrin [Geosigma AB, Uppsala (Sweden); Tullborg, Eva-Lena [Terralogica AB, Graabo (Sweden); Smellie, John [Conterra AB, Uppsala (Sweden); Gimeno, Maria J.; Gomez, Javier B.; Auque, Luis F. [Univ. of Zaragoza, Zaragoza (Spain); Sandstroem, Bjoern [WSP Sverige AB, Goeteborg (Sweden); Pedersen, Karsten [Micans AB, Moelnlycke (Sweden)

    2011-11-15

    There are plans that the final repository for low and intermediate level radioactive waste, SFR, located about 150 km north of Stockholm, will be extended. Geoscientific studies to define and characterise a suitable bedrock volume for the extended repository have been carried out from 2007 to 2011, and have included the drilling and evaluation of seven new core drilled and four percussion boreholes. These new data, together with existing data extending back to 1985, have been interpreted and modelled in order to provide the necessary information for safety assessment and repository design. This report presents the final hydrogeochemical site description for the SFR site, and will constitute a background report for the integrated site description (the SFR Site Descriptive Model version 1.0) together with corresponding reports from the geological and hydrogeological disciplines. Most of the hydrogeochemical data from the field investigations consist of major ions and isotopes together with sporadic gas, microbe and measured redox data. Despite the close proximity of the Forsmark site, few data from this source are of relevance because of the shallow nature of the SFR site, the fact that SFR is located beneath the Baltic Sea and also the drawdown/upconing impacts of its construction on the hydrogeochemistry. This artificially imposed dynamic flow system is naturally more prevalent along major deformation fracture zones of higher transmissivity, whilst lower transmissive fractures together with the less transmissive bedrock masses between major deformation zones, still retain some evidence of the natural groundwater mixing patterns established prior to the SFR construction. The groundwaters in the SFR dataset cover a depth down to -250 m.a.s.l. with single sampling locations at -300 and -400 m.a.s.l. and represent a relatively limited salinity range (1,500 to 5,500 mg/L chloride). However, the {delta}{sup 18}O values show a wide variation (-15.5 to -7.5 per mille V

  7. Fabrication of preliminary fuel rods for SFR

    International Nuclear Information System (INIS)

    Kim, Sun Ki; Oh, Seok Jin; Ko, Young Mo; Woo, Youn Myung; Kim, Ki Hwan

    2012-01-01

    Metal fuels was selected for fueling many of the first reactors in the US, including the Experimental Breeder Reactor-I (EBR-I) and the Experimental Breeder Reactor-II (EBR-II) in Idaho, the FERMI-I reactor, and the Dounreay Fast Reactor (DFR) in the UK. Metallic U.Pu.Zr alloys were the reference fuel for the US Integral Fast Reactor (IFR) program. Metallic fuel has advantages such as simple fabrication procedures, good neutron economy, high thermal conductivity, excellent compatibility with a Na coolant and inherent passive safety. U-Zr-Pu alloy fuels have been used for SFR (sodium-cooled fast reactor) related to the closed fuel cycle for managing minor actinides and reducing a high radioactivity levels since the 1980s. Fabrication technology of metallic fuel for SFR has been in development in Korea as a national nuclear R and D program since 2007. For the final goal of SFR fuel rod fabrication with good performance, recently, three preliminary fuel rods were fabricated. In this paper, the preliminary fuel rods were fabricated, and then the inspection for QC(quality control) of the fuel rods was performed

  8. The progress and efficiency on SFR

    International Nuclear Information System (INIS)

    Li Shisen.

    1985-01-01

    The study comprehends an analysis of construction management at the work site of SFR, the final repository for low and medium level nuclear waste, situated in Forsmark. The period of analysis is 1985. During this year, the most intensive part of the rock excavation work of totaly 430,000 solid cubic metres took place. Many tunnels and big chambers as well as a huge silo were driven. Many drawings and figures are given to show how the project was going on and how the construction efficiency was during the year of 1985. SFR is a highly mechanized underground project. Based on the analysis of the composition of the construction costs in SFR, the work study points out that the machine costs is the biggest part in the construction costs, and that there is a close relation between the construction costs and machine management. Some experience in reducing construction costs and some experience in construction management are introduced in detail. In order to improve the machine management, attention should be paid to increase the utilization ratio of machines. A preliminary study of time utilization ratio of drill jumbo and bolting rig is given. (orig./HP)

  9. Low and intermediate level waste in SFR-1. Reference Waste Inventory 2007

    International Nuclear Information System (INIS)

    Almkvist, Lisa; Gordon, Ann

    2007-11-01

    The objective with this report is to describe all the waste and the waste package that is expected to be deposited in SFR 1 at the time of closure. The report will form the basis for the release calculation in the safety analysis for SFR 1. Three different scenarios are explored in this report; the waste inventory is based on an estimated operational lifetime of the Swedish nuclear power plants of 50 and 60 years and that closure of the SFR 1 repository will take place in 2040 or 2050 respectively. The third scenario is where the repository is full (one part where the activity adds up to 1016 Bq and one part where the repository is considered full regarding volume). In the report, data about geometries, weights, materials, chemicals and radionuclide are given. No chemotoxic material has been identified in the waste. The inventory is estimated using the Prosit-interface which extracts information from the Triumf database. The inventory is based on so called 'waste types' and the waste types' 'reference waste package'. The reference waste package combined with a prognosis of the number of waste packages to be delivered to SFR 1 gives the final waste inventory for SFR 1. All reference waste packages are thoroughly described in the appendices of this report. The reference waste packages are as far as possible based on actual experiences and measurements. The radionuclide inventory is also based on actual measurements. The inventory is based on measurements of 60 Co and 137 Cs in waste packages and on measurements of 239 Pu and 240 Pu in reactor water. Other nuclides in the inventory are calculated with correlation factors

  10. Low and intermediate level waste in SFR-1. Reference Waste Inventory 2007

    Energy Technology Data Exchange (ETDEWEB)

    Almkvist, Lisa (Vattenfall Power Consultant AB, Stockholm (SE)); Gordon, Anna (Swedish Nuclear Fuel and Waste Management Co., Stockholm (SE))

    2007-11-15

    The objective with this report is to describe all the waste and the waste package that is expected to be deposited in SFR 1 at the time of closure. The report will form the basis for the release calculation in the safety analysis for SFR 1. Three different scenarios are explored in this report; the waste inventory is based on an estimated operational lifetime of the Swedish nuclear power plants of 50 and 60 years and that closure of the SFR 1 repository will take place in 2040 or 2050 respectively. The third scenario is where the repository is full (one part where the activity adds up to 1016 Bq and one part where the repository is considered full regarding volume). In the report, data about geometries, weights, materials, chemicals and radionuclide are given. No chemotoxic material has been identified in the waste. The inventory is estimated using the Prosit-interface which extracts information from the Triumf database. The inventory is based on so called 'waste types' and the waste types' 'reference waste package'. The reference waste package combined with a prognosis of the number of waste packages to be delivered to SFR 1 gives the final waste inventory for SFR 1. All reference waste packages are thoroughly described in the appendices of this report. The reference waste packages are as far as possible based on actual experiences and measurements. The radionuclide inventory is also based on actual measurements. The inventory is based on measurements of 60Co and 137Cs in waste packages and on measurements of 239Pu and 240Pu in reactor water. Other nuclides in the inventory are calculated with correlation factors

  11. Evaluation of alternative fluids for SFR intermediate loops

    International Nuclear Information System (INIS)

    Brissonneau, L.; Simon, N.; Baque, F.

    2009-01-01

    Among the Generation IV systems, Sodium Fast Reactors (SFR) are promising and benefit of considerable technological experience, but improvements are researched on safety approach and capital cost reduction. One of the main drawback to be solved by the standard SFR design is the proper management of the risk of leakage between the intermediate circuit filled with sodium and the energy conversion system using a water Rankine cycle. The limitation of this risk requires notably an early detection of water leakage to prevent a water-sodium reaction. One innovative solution consists in the replacement of the sodium in the secondary loops by an alternative liquid fluid, not or less reactive with water. This alternative fluid might also allow innovative designs, e.g. intermediate heat exchanger and steam generator grouped in the same component. CEA, Areva NP and EdF have joined in a working group in order to evaluate different 'alternative fluids' that might replace sodium. A first selection retained seven fluids on the basis of 'required properties' as large operating range (low melting point, high boiling point ...), fluid cost and availability, acceptable corrosion at SFR working temperature. These are three bismuth alloys, two nitrate salts, one hydroxide melt and sodium with nanoparticles of nickel. Then, it was decided to evaluate these fluids through a multi-criteria analysis in order to quantify advantages and drawbacks of each fluid and to compare them with sodium. Lack of knowledge, impact on materials, design, working conditions and reactor availability should be emphasized by this analysis, in order to provide sound arguments for a research program on one or two promising fluids. A global note is given to each fluid by evaluating them with respect to 'grand criteria', weighted differently according to their importance. The grand criteria are : thermal properties, reactivity with structures, reactivity with other fluids (air, water, sodium), chemistry control

  12. Uncertainty Evaluation of the SFR Subchannel Thermal-Hydraulic Modeling Using a Hot Channel Factors Analysis

    International Nuclear Information System (INIS)

    Choi, Sun Rock; Cho, Chung Ho; Kim, Sang Ji

    2011-01-01

    In an SFR core analysis, a hot channel factors (HCF) method is most commonly used to evaluate uncertainty. It was employed to the early design such as the CRBRP and IFR. In other ways, the improved thermal design procedure (ITDP) is able to calculate the overall uncertainty based on the Root Sum Square technique and sensitivity analyses of each design parameters. The Monte Carlo method (MCM) is also employed to estimate the uncertainties. In this method, all the input uncertainties are randomly sampled according to their probability density functions and the resulting distribution for the output quantity is analyzed. Since an uncertainty analysis is basically calculated from the temperature distribution in a subassembly, the core thermal-hydraulic modeling greatly affects the resulting uncertainty. At KAERI, the SLTHEN and MATRA-LMR codes have been utilized to analyze the SFR core thermal-hydraulics. The SLTHEN (steady-state LMR core thermal hydraulics analysis code based on the ENERGY model) code is a modified version of the SUPERENERGY2 code, which conducts a multi-assembly, steady state calculation based on a simplified ENERGY model. The detailed subchannel analysis code MATRA-LMR (Multichannel Analyzer for Steady-State and Transients in Rod Arrays for Liquid Metal Reactors), an LMR version of MATRA, was also developed specifically for the SFR core thermal-hydraulic analysis. This paper describes comparative studies for core thermal-hydraulic models. The subchannel analysis and a hot channel factors based uncertainty evaluation system is established to estimate the core thermofluidic uncertainties using the MATRA-LMR code and the results are compared to those of the SLTHEN code

  13. Nuclear data propagation with burnup. Impact on SFR reactivity coefficients

    International Nuclear Information System (INIS)

    Buiron, Laurent; Plisson-Rieunier, Daniele

    2017-01-01

    For the next generation fast reactor design, the Generation IV International Forum (GIF) defined global objectives in terms of safety improvement, sustainability, waste minimization and non-proliferation. Among the possibilities studied at CEA, Sodium cooled Fast Reactor (SFR) are studied as potential industrial tools for next decade's deployment. Many efforts have been made in the last years to obtain advanced industrial core designs that comply with these goals. Concerning safety issues, particular efforts have been made in order to obtain core designs that can be resilient to accidental transients. The 'safety' level of such new designs is often characterized by their 'natural' behavior under unprotected transients such as loss of flow or hypothetical transient over power. Transient analysis needs several accurate neutronic input data such as reactivity coefficient and kinetic parameters. Beside estimation of the level of 'absolute' values, associated uncertainties have also to be evaluated for the whole set of relevant data. These estimations have to be performed for different core state such as end of cycle core for feedback coefficient. This means that uncertainties have to be obtained not only a fixed time but also have to be propagated all through irradiation. To do so, we need to couple Boltzman and Bateman equations at sensitivities level. The coupling process could be done with the help of the perturbation theory which gives adapted framework suited for deterministic calculation codes. This coupling is currently in progress in ERANOS code system. The actual implementation gives access to estimation of sensitivities for both reactivity coefficients and mass balance. After a brief theoretical description of Boltzman/Bateman coupling capabilities in ERANOS, the study presented in this paper focuses on sensitivity and uncertainties estimation for the main feedback coefficients involved in fast reactor transients: the

  14. Establishment of Collaboration System for SFR Technology Development between Korea and France

    International Nuclear Information System (INIS)

    Yoo, Jae Woon; Kim, Yeong Il; Choi, Jong Hyeun; Seong, Seung Hwan; Eoh, Jae Hyuk; Jeong, Hae Yong; Hahn, Do Hee; Lee, Yong Bum; Chang, Jin Wook; Lee, Dong Uk

    2010-03-01

    1) Review of the technical status and plane associated with STCs on the SFR R and D · The objective of the study was accomplished by constructing an human network and investigating the status on the following 5 STCs - Analysis of BFS -73 -1 and 75-1 critical experiments using ERANOS. - SFR : Elector-magnetic pumps or mechanical pumps; criteria for selection, description and modeling - Phenix end of life tests - SC-CO2 Brayton cycle : Investigation of sodium-carbon dioxide interactions; potential consequences on reactor operation - Evaluation of lead-bismuth eutectic (LBE) coolant for SFR intermediate loop 2) Holding KAERI-CEA SFR technical meeting on STCs - Final investigation at SFR technical meeting held at CEA Cadarache from Jan. 5 to Jan. 7 in 2010. - Agreement on further action plan for completing the STC - Deduction of future collaboration topics and agreed to submit into the next JCCNE - Agreed to hold next SFR technical meeting in Korea on around October 2010

  15. Security-by-design approach of the KALIMER 600 SFR plant

    International Nuclear Information System (INIS)

    So, Dong Sup; Lee, Yong Bum

    2012-01-01

    Security measures as well as safety and safeguards measures should be incorporated and addressed early in the design process to enhance the cost effectiveness of a PPS (Physical Protection System). Safety, security, operations, and safeguards design teams and regulators need to be flexible and perform 'trade studies' on the available options. In this paper, SBD (Security by Design) measures in the design phase of the KALIMER 600 SFR (Sodium Cooled Reactor) plant are identified and discussed qualitatively

  16. Bedrock Hydrogeology-Groundwater flow modelling. Site investigation SFR

    International Nuclear Information System (INIS)

    Oehman, Johan; Follin, Sven; Oden, Magnus

    2013-05-01

    The hydrogeological model developed for the SFR extension project (PSU) consists of 40 geologically modelled deformation zones (DZ) and 8 sub-horizontal structural-hydraulic features, called SBAstructures, not defined in the geological model. However, some of the SBA-structures coincide with what is defined as unresolved possible deformation zones (Unresolved PDZ) in the geological modelling. In addition, the hydrogeological model consists of a stochastic discrete fracture network (DFN) model intended for the less fractured rock mass volumes (fracture domains) between the zones and the SBA-structures, and a stochastic fracture model intended to handle remaining Unresolved PDZs in the geological modelling not modelled as SBA-structures in the hydrogeological modelling. The four structural components of the bedrock in the hydrogeological model, i.e. DZ, SBA, Unresolved PDZ and DFN, are assigned hydraulic properties in the hydrogeological model based on the transmissivities interpreted from single-hole hydraulic tests. The main objective of the present work is to present the characteristics of the hydrogeological model with regard to the needs of the forthcoming safety assessment SR-PSU. In concrete words, simulated data are compared with measured data, i.e. hydraulic heads in boreholes and tunnel inflow to the existing repository (SFR). The calculations suggest that the available data for flow model calibration cannot be used to motivate a substantial adjustment of the initial hydraulic parameterisation (assignment of hydraulic properties) of the hydrogeological model. It is suggested that uncertainties in the hydrogeological model are studied in the safety assessment SR-PSU by means of a large number of calculation cases. These should address hydraulic heterogeneity of deterministic structures (DZ and SBA) and realisations of stochastic fractures/fracture networks (Unresolved PDZ and DFN) within the entire SFR Regional model domain

  17. Bedrock Hydrogeology - Groundwater flow modelling. Site investigation SFR

    Energy Technology Data Exchange (ETDEWEB)

    Oehman, Johan [Geosigma AB, Uppsala (Sweden); Follin, Sven [SF GeoLogic AB, Taeby (Sweden); Oden, Magnus [SKB, Stockholm (Sweden)

    2013-05-15

    The hydrogeological model developed for the SFR extension project (PSU) consists of 40 geologically modelled deformation zones (DZ) and 8 sub-horizontal structural-hydraulic features, called SBAstructures, not defined in the geological model. However, some of the SBA-structures coincide with what is defined as unresolved possible deformation zones (Unresolved PDZ) in the geological modelling. In addition, the hydrogeological model consists of a stochastic discrete fracture network (DFN) model intended for the less fractured rock mass volumes (fracture domains) between the zones and the SBA-structures, and a stochastic fracture model intended to handle remaining Unresolved PDZs in the geological modelling not modelled as SBA-structures in the hydrogeological modelling. The four structural components of the bedrock in the hydrogeological model, i.e. DZ, SBA, Unresolved PDZ and DFN, are assigned hydraulic properties in the hydrogeological model based on the transmissivities interpreted from single-hole hydraulic tests. The main objective of the present work is to present the characteristics of the hydrogeological model with regard to the needs of the forthcoming safety assessment SR-PSU. In concrete words, simulated data are compared with measured data, i.e. hydraulic heads in boreholes and tunnel inflow to the existing repository (SFR). The calculations suggest that the available data for flow model calibration cannot be used to motivate a substantial adjustment of the initial hydraulic parameterisation (assignment of hydraulic properties) of the hydrogeological model. It is suggested that uncertainties in the hydrogeological model are studied in the safety assessment SR-PSU by means of a large number of calculation cases. These should address hydraulic heterogeneity of deterministic structures (DZ and SBA) and realisations of stochastic fractures/fracture networks (Unresolved PDZ and DFN) within the entire SFR Regional model domain.

  18. A carbon budget for the aquatic ecosystem above SFR in Oeregrundsgrepen

    Energy Technology Data Exchange (ETDEWEB)

    Kumblad, L [Stockholm Univ. (Sweden). Dept. of Systems Ecology

    1999-07-01

    The potential hazards of radionuclide release to humans and the environment is regularly evaluated in safety assessments of SFR, the final repository for radioactive operational waste. SFR handles, since 1988, low and intermediate level nuclear waste from Swedish nuclear power plants, medical care attendance, industries and research laboratories and is located in the bedrock 50 meters under the seabed of Oeregrundsgrepen in the southern Bothnian Sea. This report presents a description of the aquatic ecosystem and a carbon budget for the area above SFR with the aim to include ecosystem dynamics in the present safety assessment of the repository (SAFE). The carbon budget will support SAFE by facilitating evaluations of transport and fate of radionuclides, primarily {sup 14}C, in case of a release from the repository and describe the ecosystem structure and function. Furthermore, {sup 14}C is the dose-dominant radionuclide in the repository which most likely will follow the general carbon flow in the ecosystem if there should be a release. The carbon budget was based on biomass and flow of carbon between thirteen functional groups (including POC and DOC) in the ecosystem above SFR and the results indicates that the organisms are self-sufficient on carbon and that the area exports carbon corresponding to approximately 50% of the annual primary production. The largest organic carbon pool is DOC (one and a half time larger than the total biomass) and the major functional organism groups are the macrophytes (37% of the total biomass), benthic macrofauna (36%), and the microphytes (11%). The soft bottom and phytobenthic communities seem to have important roles in the ecosystem since these communities comprise the main part of the living carbon in the studied area.

  19. A carbon budget for the aquatic ecosystem above SFR in Oeregrundsgrepen

    International Nuclear Information System (INIS)

    Kumblad, L

    1999-07-01

    The potential hazards of radionuclide release to humans and the environment is regularly evaluated in safety assessments of SFR, the final repository for radioactive operational waste. SFR handles, since 1988, low and intermediate level nuclear waste from Swedish nuclear power plants, medical care attendance, industries and research laboratories and is located in the bedrock 50 meters under the seabed of Oeregrundsgrepen in the southern Bothnian Sea. This report presents a description of the aquatic ecosystem and a carbon budget for the area above SFR with the aim to include ecosystem dynamics in the present safety assessment of the repository (SAFE). The carbon budget will support SAFE by facilitating evaluations of transport and fate of radionuclides, primarily 14 C, in case of a release from the repository and describe the ecosystem structure and function. Furthermore, 14 C is the dose-dominant radionuclide in the repository which most likely will follow the general carbon flow in the ecosystem if there should be a release. The carbon budget was based on biomass and flow of carbon between thirteen functional groups (including POC and DOC) in the ecosystem above SFR and the results indicates that the organisms are self-sufficient on carbon and that the area exports carbon corresponding to approximately 50% of the annual primary production. The largest organic carbon pool is DOC (one and a half time larger than the total biomass) and the major functional organism groups are the macrophytes (37% of the total biomass), benthic macrofauna (36%), and the microphytes (11%). The soft bottom and phytobenthic communities seem to have important roles in the ecosystem since these communities comprise the main part of the living carbon in the studied area

  20. Analysis of solutions for passively activated safety shutdown devices for SFR

    International Nuclear Information System (INIS)

    Burgazzi, Luciano

    2013-01-01

    Highlights: • Innovative systems for emergency shut down of fast reactors are proposed. • The concepts of inherent and passive safety are put forward. • The relative analysis in terms of safety and reliability is presented. • A comparative assessment among the concepts is performed. • Path forward is tracked. -- Abstract: In order to enhance the inherent safety of fast reactors, innovative reactivity control systems have been proposed for intrinsic ultimate shut-down instead of conventional scram rods, to cope with the potential consequences of severe unprotected transient accidents, such as an energetic core disruptive accident, as in case of sodium fast reactors. The passive shut-down systems are designed to shut-down system only by inherent passive reactivity feedback mechanism, under unprotected accident conditions, implying failure of reactor protection system. They are conceived to be self-actuated without any signal elaboration, since the actuation of the system is triggered by the effects induced by the transient like material dilatation, in case of overheating of the coolant for instance, according to fast reactor design to meet the safety requirements. This article looks at different special shutdown systems specifically engineered for prevention of severe accidents, to be implemented on fast reactors, with main focus on the investigation of the performance of the self-actuated shutdown systems in sodium fast reactors

  1. Feasibility Study on Two-phase Thermosiphon for External Vessel Cooling Application of SFR

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Jae Young; Song, Sub Lee; Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2014-05-15

    This study shows that ex-vessel cooling by two-phase thermosiphon is feasible for large size of SFR. The result presents that further studies to increase heat transfer on condenser-air and gap is necessary and the experiment should be conducted for the validation. Also, the heat loss through evaporator during normal operation, corrosion, consideration of organic fluid to exclude the poison of mercury should be studied. As the necessity of sodium fast reactor in order to reduce spent fuel, the development of designing sodium fast reactor becomes an issue. Even though there is PDRC and RVACS for the decay heat removal (DHR) system, each system has disadvantage of sodium fire and low performance, respectively. Therefore, to increase the safety of SFR, the new passive safety system design is needed without sodium fire and high performance, which can applied for large SFR. The DHR system using two-phase thermosiphon for external vessel cooling application is suggested in this paper. The proposed design have advantage that there is no structure in reactor vessel, which means no system modification and no sodium fire with perfect isolation. Also, it provide the method to mitigate sodium fire in case of sodium leakage from reactor vessel.

  2. Core Thermal-Hydraulic Conceptual Design for the Advanced SFR Design Concepts

    International Nuclear Information System (INIS)

    Cho, Chung Ho; Chang, Jin Wook; Yoo, Jae Woon; Song, Hoon; Choi, Sun Rock; Park, Won Seok; Kim, Sang Ji

    2010-01-01

    The Korea Atomic Energy Research Institute (KAERI) has developed the advanced SFR design concepts from 2007 to 2009 under the National longterm Nuclear R and D Program. Two types of core designs, 1,200 MWe breakeven and 600 MWe TRU burner core have been proposed and evaluated whether they meet the design requirements for the Gen IV technology goals of sustainability, safety and reliability, economics, proliferation resistance, and physical protection. In generally, the core thermal hydraulic design is performed during the conceptual design phase to efficiently extract the core thermal power by distributing the appropriate sodium coolant flow according to the power of each assembly because the conventional SFR core is composed of hundreds of ducted assemblies with hundreds of fuel rods. In carrying out the thermal and hydraulic design, special attention has to be paid to several performance parameters in order to assure proper performance and safety of fuel and core; the coolant boiling, fuel melting, structural integrity of the components, fuel-cladding eutectic melting, etc. The overall conceptual design procedure for core thermal and hydraulic conceptual design, i.e., flow grouping and peak pin temperature calculations, pressure drop calculations, steady-state and detailed sub-channel analysis is shown Figure 1. In the conceptual design phase, results of core thermal-hydraulic design for advanced design concepts, the core flow grouping, peak pin cladding mid-wall temperature, and pressure drop calculations, are summarized in this study

  3. Level II Probabilistic Safety Analysis Methodology for the Application to GEN-IV Sodium-cooled Fast Reactor

    International Nuclear Information System (INIS)

    Park, S. Y.; Kim, T. W.; Han, S. H.; Jeong, H. Y.

    2010-03-01

    The Korea Atomic Energy Research Institute (KAERI) has been developing liquid metal reactor (LMR) design technologies under a National Nuclear R and D Program. Nevertheless, there is no experience of the probabilistic safety assessment (PSA) domestically for a fast reactor with the metal fuel. Therefore, the objective of this study is to establish the methodologies of risk assessment for the reference design of GEN-IV sodium fast reactor (SFR). An applicability of the PSA methodology of U. S. NRC and PRISM plant to the domestic GEN-IV SFR has been studied. The study contains a plant damage state analysis, a containment event tree analysis, and a source-term release category binning process

  4. Safety Design Criteria (SDC) for Gen-IV Sodium-cooled Fast Reactor

    International Nuclear Information System (INIS)

    Nakai, Ryodai

    2013-01-01

    SDC Development Background & Objectives: • Safety Design Criteria (SDC) Development for Gen-IV SFR: – Proposed at the GIF Policy Group (PG) meeting in October 2010 –SDC “harmonization” is increasingly important for: • Realization of enhanced safety designs meeting to Gen-IV safety goals and safety approach common to SFR systems; • Preparation for the forthcoming licensing in the near future; • Because Gen-IV SFR are progressing into conceptual design stage. • The SDC is the Reference criteria: – Of the designs of safety-related Structures, Systems & Components that are specific to the SFR system; – For clarifying the requisites systematically & comprehensively; – When the technology developers apply the basic safety approach and use the codes & standards for conceptual design of the Gen-IV SFR system

  5. The nuclide inventory in SFR-1

    International Nuclear Information System (INIS)

    Ingemansson, Tor

    2001-10-01

    This report is an account for a project carried out on behalf of the Swedish Radiation Protection Authority (SSI): 'Nuclide inventory in SFR-1' (The Swedish underground disposal facility for low and intermediate level reactor waste). The project comprises the following five sub-projects: 1) Measuring methods for nuclides, difficult to measure, 2) The nuclide inventory in SFR-1, 3) Proposal for nuclide library for SFR-1 and ground disposal, 4) Nuclide library for exemption, and 5) Characterising of the nuclide inventory and documentation for SFL waste. In all five sub-projects long-lived activity, including Cl-36, has been considered

  6. One-Dimensional Analysis of Thermal Stratification in AHTR and SFR Coolant Pools

    International Nuclear Information System (INIS)

    Haihua Zhao; Per F. Peterson

    2007-01-01

    Thermal stratification phenomena are very common in pool type reactor systems, such as the liquid-salt cooled Advanced High Temperature Reactor (AHTR) and liquid-metal cooled fast reactor systems such as the Sodium Fast Reactor (SFR). It is important to accurately predict the temperature and density distributions both for design optimation and accident analysis. Current major reactor system analysis codes such as RELAP5 (for LWR's, and recently extended to analyze high temperature reactors), TRAC (for LWR's), and SASSYS (for liquid metal fast reactors) only provide lumped-volume based models which can only give very approximate results and can only handle simple cases with one mixing source. While 2-D or 3-D CFD methods can be used to analyze simple configurations, these methods require very fine grid resolution to resolve thin substructures such as jets and wall boundaries, yet such fine grid resolution is difficult or impossible to provide for studying the reactor response to transients due to computational expense. Therefore, new methods are needed to support design optimization and safety analysis of Generation IV pool type reactor systems. Previous scaling has shown that stratified mixing processes in large stably stratified enclosures can be described using one-dimensional differential equations, with the vertical transport by free and wall jets modeled using standard integral techniques. This allows very large reductions in computational effort compared to three-dimensional numerical modeling of turbulent mixing in large enclosures. The BMIX++ (Berkeley mechanistic MIXing code in C++) code was originally developed at UC Berkeley to implement such ideas. This code solves mixing and heat transfer problems in stably stratified enclosures. The code uses a Lagrangian approach to solve 1-D transient governing equations for the ambient fluid and uses analytical or 1-D integral models to compute substructures. By including liquid salt properties, BMIX++ code is

  7. Low and intermediate level waste in SFR-1. Reference waste inventory

    International Nuclear Information System (INIS)

    Riggare, P.; Johansson, Claes

    2001-06-01

    The objective with this report is to describe all the waste and the waste package that is expected to be deposited in SFR-1 at the time of closure. This report is a part of the SAFE project (Safety Assessment of Final Repository for Radioactive Operational Waste), i.e. the renewed safety assessment of SFR-1. The accounted waste inventory has been used as input to the release calculation that has been performed in the SAFE project. The waste inventory is based on an estimated operational lifetime of the Swedish nuclear power plants of 40 years and that closure of the SFR repository will happen in 2030. In the report, data about geometries, weights, materials, chemicals and radionuclide are given. No chemo toxic material has been identified in the waste. The inventory is based on so called waste types and the waste types reference waste package. The reference waste package combined with a prognosis of the number of waste packages to the year 2030 gives the final waste inventory for SFR-1. All reference waste packages are thoroughly described in the appendices of this report. The reference waste packages are as far as possible based on actual experiences and measurements. The radionuclide inventory is also based on actual measurements. The inventory is based on measurements of 60 Co and 137 Cs in waste packages and on measurements 239 Pu and 240 Pu in reactor water. Other nuclides in the inventory are calculated with correlation factors. In the SAFE project's prerequisites it was said that one realistic and one conservative (pessimistic) inventory should be produced. The conservative one should then be used for the release calculations. In this report one realistic and one conservative radionuclide inventory is presented. The conservative one adds up to 10 16 Bq. Regarding materials there is only one inventory given since it is not certain what is a conservative assumption

  8. Project SAFE. Update of the SFR-1 safety assessment. Phase 1. Appendix A2: Scenarios

    International Nuclear Information System (INIS)

    Skagius, K.; Wiborgh, M.

    1998-01-01

    This appendix gives a short description of the scenario methodology adopted in the previous safety assessment of SFR. Since then new methodologies for developing structured descriptions of how processes and interactions between processes affect the evolution of a repository system. Two such methods are briefly described. These methods are very similar, but they differ in the way the system is graphically structured. One of the methods is based on Process Influence Diagrams, PID, and the other on Interaction matrices. It is proposed that the method based on Interaction matrices is used for the scenario work in project SAFE. The main reason for this is that the method already has been applied by SKB, which means that it will be possible to use already existing procedures and documentation systems. The proposed procedure involves the development of Interaction matrices for a defined Reference scenario and the use of these matrices to illustrate the effect of different Scenario initiating FEPs. The proposed procedure is described in this appendix

  9. Preliminary Hydrogeochemical Site Description SFR (version 0.2)

    Energy Technology Data Exchange (ETDEWEB)

    Nilsson, Ann-Chatrin (Geosigma AB, Uppaala (Sweden)); Tullborg, Eva-Lena (Terralogica AB, Graabo (Sweden)); Smellie, John (Conterra AB, Partille (Sweden))

    2010-05-15

    The final repository for low and intermediate level radioactive operational waste, SFR, located about 150 km north of Stockholm, is to undergo a future extension. The present on-going project, scheduled from 2007 to 2011, is to define and characterise a suitable bedrock volume for the extended repository. This will include the drilling and geoscientific evaluation of seven core-drilled and four percussion boreholes as well as subsequent interpretation and modelling based on the obtained results in order to provide the necessary information for safety assessment and repository design. This report presents a preliminary hydrogeochemical site description for the SFR site and should be considered as an early progress report rather than a complete hydrochemical site descriptive model. The completed hydrogeochemical field investigations have yielded chemical data from a total of 12 borehole sections in five boreholes and additional data from the entire length of two open boreholes in connection with hydraulic tests. These data, together with data from a total of 18 early boreholes in the present SFR tunnel system, were used in the interpretation work. The main part of the data consisted of basic groundwater analyses including major ions and isotopes. Some sporadic gas, microbe and measured redox data are available, but these are either not treated in this report, or are only briefly discussed. This was due to time constraints since special care is needed when interpreting few data of varying quality. The groundwaters in the SFR dataset cover a maximum depth down to about .400 masl and represent a relatively limited salinity range (1,500 to 5,500 mg/L chloride). However, the delta18O values show a wide variation (-1.55 to -0.75% V-SMOW) similar to that reported from the Forsmark site investigations. At the SFR, marine indicators such as Mg/Cl, K/Cl and Br/Cl also show relatively large variations considering the limited salinity range. From very few measured Eh values, and

  10. Preliminary Hydrogeochemical Site Description SFR (version 0.2)

    International Nuclear Information System (INIS)

    Nilsson, Ann-Chatrin; Tullborg, Eva-Lena; Smellie, John

    2010-05-01

    The final repository for low and intermediate level radioactive operational waste, SFR, located about 150 km north of Stockholm, is to undergo a future extension. The present on-going project, scheduled from 2007 to 2011, is to define and characterise a suitable bedrock volume for the extended repository. This will include the drilling and geoscientific evaluation of seven core-drilled and four percussion boreholes as well as subsequent interpretation and modelling based on the obtained results in order to provide the necessary information for safety assessment and repository design. This report presents a preliminary hydrogeochemical site description for the SFR site and should be considered as an early progress report rather than a complete hydrochemical site descriptive model. The completed hydrogeochemical field investigations have yielded chemical data from a total of 12 borehole sections in five boreholes and additional data from the entire length of two open boreholes in connection with hydraulic tests. These data, together with data from a total of 18 early boreholes in the present SFR tunnel system, were used in the interpretation work. The main part of the data consisted of basic groundwater analyses including major ions and isotopes. Some sporadic gas, microbe and measured redox data are available, but these are either not treated in this report, or are only briefly discussed. This was due to time constraints since special care is needed when interpreting few data of varying quality. The groundwaters in the SFR dataset cover a maximum depth down to about .400 masl and represent a relatively limited salinity range (1,500 to 5,500 mg/L chloride). However, the δ 18 O values show a wide variation (-1.55 to -0.75% V-SMOW) similar to that reported from the Forsmark site investigations. At the SFR, marine indicators such as Mg/Cl, K/Cl and Br/Cl also show relatively large variations considering the limited salinity range. From very few measured Eh values, and

  11. Experimental Facilities and Plan for a Prototype SFR

    International Nuclear Information System (INIS)

    Chang, Jinwook

    2013-01-01

    Contents of the presentation: 1. STELLA; 2. Under Sodium Viewing; 3. Sodium-CO 2 Interaction Test; Overview of the Sodium Integral Effect Test Loop for Safety Simulation and Assessment (STELLA) Program, the scope of the experiment and the overall characteristics of STELLA-1: Phase 1: STELLA-1, Individual component test; • Performance evaluation of key sodium components; • Heat exchanger design codes V&V. Phase 2: STELLA-2, Integral effect test; • Verification of dynamic plant response after reactor shutdown; • Construction of test DB to support specific design approval for the prototype SFR

  12. Numerical investigation on turbulent natural convection in partially connected cylindrical enclosures for analysing SFR safety under core meltdown scenario

    International Nuclear Information System (INIS)

    David, Dijo K.; Mangarjuna Rao, P.; Nashine, B.K.; Selvaraj, P.

    2015-01-01

    Under the unlikely event of severe core meltdown accident in pool type SFR, the molten core materials may rupture the grid plate which supports the fuel subassemblies and it can get relocated in to the lower pool. These debris may eventually settle on the debris collector (i.e., core catcher) installed above the bottom wall of the lower pool. The bed thus formed generates heat due to radioactive decay which has to be passively removed for maintaining the structural integrity of main vessel. By means of natural convection, the heat generated in the debris bed will be transferred to the top pool where the heat sink (i.e., Decay heat exchanger (DHX)) is installed. Heat transfer to the DHX (which is a part of safety grade decay heat removal system) can take place through the opening created in the grid plate which connects the two liquid pools (i.e., the top pool and the lower pool). Heat transfer can also take place through the lateral wall of the lower cylindrical pool to the side pool and eventually to the top pool, and thus to the DHX. This study numerically investigates the effectiveness of heat transfer between lower pool and top pool during PARR by considering them as partially connected cylindrical enclosures. The governing equations have been numerically solved using finite volume method in cylindrical co-ordinates using SIMPLE algorithm. Turbulence has been modeled using k-ω model and the model is validated against benchmark problems of natural convection found in literature. The effect of parameters such as the heat generation rate in the bed and the size of the grid plate opening are evaluated. Also PAHR in SFR pool is modeled using an axi-symmetric model to fund out the influence of grid plate opening on heat removal from core catcher. The results obtained are useful for improving the cooling capability of in-vessel tray type core catcher for handling the whole core meltdown scenarios in SFR. (author)

  13. SFR Safety Consideration in Light of Fukushima Dai-ichi Accident

    International Nuclear Information System (INIS)

    Yamaguchi, Akira

    2013-01-01

    SFR Considerations: Fukushima Dai-ichi Accident: • Combined LORL and LOHS type initiated from SBO; • High pressure water-steam cooling system: – Depressurization - Not needed; – Ultimate heat sink - Robust (NC to atmosphere); – Continuous injection - Not needed (large sensible heat capacity). • Severe accident management: – RPV failure resulted in depressurization - Elevated temperature; – Heat sink to atmosphere - Freeing risk, sodium fire risk; – Mobile power supply - External resource may not be needed; – Seawater injection with fire engines - Sodium injection not needed; • Containment performance and accessibility: – Containment - Large containment volume and low pressure system; – Explosives - Sodium fire and hydrogen explosion

  14. Evaluation of Spent Fuel Recycling Scenario using Pyro-SFR related System

    International Nuclear Information System (INIS)

    Lee, Yong Kyo; Kim, Sang Ji; Kim, Young Jin

    2014-01-01

    It is needed to validate whether the recycling scenario connecting pyro-processing and sodium-cooled fast reactor(SFR) is promising or not. The latest technologies of pyro-processing are applied to SFR and the recycling scenario is evaluated through the SFR's performance analysis. The analyzed SFR is KALIMER-600 TRU burner which purpose is to transmute transuranics (TRU). National policy of CANDU SF management has not been decided yet. However, the stored quantity of this SF is large enough not to be neglected. So this study includes additionally the recycling scenario of CANDU SF. Adopting the mass ratio of TRU and RE recovered in pyro-processing is 4 to 1 on PWR SF recycling, the sodium void reactivity is higher than design basis of metal fuel. So the current pyro-processing technology is may not be acceptable. If pyro-processing technology of CANDU SF is assumed to be the same as PWR's case, CANDU recycling scenario is acceptable. Transmutation performance is worse than PWR's, while the sodium void reactivity is within design limit

  15. SFR inverse modelling Part 2. Uncertainty factors of predicted flow in deposition tunnels and uncertainty in distribution of flow paths from deposition tunnels

    International Nuclear Information System (INIS)

    Holmen, Johan

    2007-10-01

    The Swedish Nuclear Fuel and Waste Management Co (SKB) is operating the SFR repository for low- and intermediate-level nuclear waste. An update of the safety analysis of SFR was carried out by SKB as the SAFE project (Safety Assessment of Final Disposal of Operational Radioactive Waste). The aim of the project was to update the safety analysis and to produce a safety report. The safety report has been submitted to the Swedish authorities. This study is a continuation of the SAFE project, and concerns the hydrogeological modelling of the SFR repository, which was carried out as part of the SAFE project, it describes the uncertainty in the tunnel flow and distributions of flow paths from the storage tunnels. Uncertainty factors are produced for two different flow situations, corresponding to 2,000 AD (the sea covers the repository) and 4,000 AD (the sea has retreated form the repository area). Uncertainty factors are produced for the different deposition tunnels. The uncertainty factors are discussed in Chapter 2 and two lists (matrix) of uncertainty factors have been delivered as a part of this study. Flow paths are produced for two different flow situations, corresponding to 2,000 AD (the sea covers the repository) and 5,000 AD (the sea has retreated form the repository area). Flow paths from the different deposition tunnels have been simulated, considering the above discussed base case and the 60 realisation that passed all tests of this base case. The flow paths are presented and discussed in Chapter 3 and files presenting the results of the flow path analyses have been delivered as part of this study. The uncertainty factors (see Chapter 2) are not independent from the flow path data (see Chapter 3). When stochastic calculations are performed by use of a transport model and the data presented in this study is used as input to such calculations, the corresponding uncertainty factors and flow path data should be used. This study also includes a brief discussion of

  16. Characterization of a sodium-cooled fast reactor in an MHR-SFR synergy for TRU transmutation

    International Nuclear Information System (INIS)

    Hong, Ser Gi; Kim, Yonghee; Venneri, Francesco

    2008-01-01

    In the task of destroying the light water reactor (LWR) transuranics (TRUs), we consider the concept of a synergistic combination of a deep-burn (DB) gas-cooled reactor followed by a sodium-cooled fast reactor (SFR), as an alternative way to the direct feeding of the LWR TRUs to the SFR. In the synergy concept, TRUs from LWR are first deeply incinerated in a graphite-moderated DB-MHR (modular helium reactor) and then the spent fuels of DB-MHR are recycled into the closed-cycle SFR. The DB-MHR core is 100% TRU-loaded and a deep-burning (50-65%) is achieved in a safe manner (as discussed in our previous work). In this analysis, the SFR fuel cycle is closed with a pyro-processing technology to minimize the waste stream to a final repository. Neutronic characteristics of the SFR core in the MHR-SFR synergy have been evaluated from the core physics point of view. Also, we have compared core characteristics of the synergy SFR with those of a stand-alone SFR transuranic burner. For a consistent comparison, the two SFRs are designed to have the same TRU consumption rate of ∼250 kg/GW EFPY that corresponds to the TRU discharge rate from three 600 MW DB-MHRs. The results of our work show that the synergy SFR, fed with TRUs from DB-MHR, has a much smaller burnup reactivity swing, a slightly greater delayed neutron fraction (both positive features) but also a higher sodium void worth and a less negative Doppler coefficients than the conventional SFR, fed with TRUs directly from the LWRs. In addition, several design measures have been considered to reduce the sodium void worth in the synergy SFR core

  17. Current status of SFR development in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Ieda, Yoshiaki; Chikazawa, Yoshitaka [Japan Atomic Energy Agency, Tokyo (Japan). Project Promotion Office; Kotake, Shoji [Japan Atomic Power Company, Tokyo (Japan)

    2012-03-15

    Fast Reactor development experiences and status in Japan are summarized. Even though international SFR circumstances were against in 1980s and 1990s, e.g. CRBRP, SNR-300 and Superphenix terminations, we kept on with our R and D activities steadily aiming at positive development targets in Japan. As results of our efforts, it has shown that our commercialized SFR concept, Japan Sodium-cooled Fast Reactor (JSFR) could meet the targets in the Feasibility Study on Commercialized Fast Reactor Cycle Systems (FS) and the Fast Reactor Cycle Technology Development (FaCT) project. Further, Monju has finally achieved restart in May 2010 after having been shut for almost 15 years. A future plan of Monju is to be determined based on a direction of the national nuclear and energy policies that will be established in 2012. The undergoing FaCT project is pursuing commercialization of fast reactor cycle system around 2050 under cooperation of MEXT (Ministry of Education, Culture, Sports, Science and Technology), METI (Ministry of Economy, Trade and Industry), utilities, venders and JAEA (Japan Atomic Energy Agency). As results of the FaCT Phase I, feasibility of the key technologies for JSFR has been evaluated and the project is waiting for launching the phase II due to the Tohoku large earthquake. It is considered that the nuclear development policy might be affected by the Tohoku large Earthquake/Tsunami in Japan. Nevertheless the significance of nuclear energy will not be changed and thus we will focus on the issues learnt from Fukushima accidents and reflect into the improvement of the safety of Monju and the safety design criteria for the next generation Fast Reactor systems. (orig.)

  18. Current status of SFR development in Japan

    International Nuclear Information System (INIS)

    Ieda, Yoshiaki; Chikazawa, Yoshitaka

    2012-01-01

    Fast Reactor development experiences and status in Japan are summarized. Even though international SFR circumstances were against in 1980s and 1990s, e.g. CRBRP, SNR-300 and Superphenix terminations, we kept on with our R and D activities steadily aiming at positive development targets in Japan. As results of our efforts, it has shown that our commercialized SFR concept, Japan Sodium-cooled Fast Reactor (JSFR) could meet the targets in the Feasibility Study on Commercialized Fast Reactor Cycle Systems (FS) and the Fast Reactor Cycle Technology Development (FaCT) project. Further, Monju has finally achieved restart in May 2010 after having been shut for almost 15 years. A future plan of Monju is to be determined based on a direction of the national nuclear and energy policies that will be established in 2012. The undergoing FaCT project is pursuing commercialization of fast reactor cycle system around 2050 under cooperation of MEXT (Ministry of Education, Culture, Sports, Science and Technology), METI (Ministry of Economy, Trade and Industry), utilities, venders and JAEA (Japan Atomic Energy Agency). As results of the FaCT Phase I, feasibility of the key technologies for JSFR has been evaluated and the project is waiting for launching the phase II due to the Tohoku large earthquake. It is considered that the nuclear development policy might be affected by the Tohoku large Earthquake/Tsunami in Japan. Nevertheless the significance of nuclear energy will not be changed and thus we will focus on the issues learnt from Fukushima accidents and reflect into the improvement of the safety of Monju and the safety design criteria for the next generation Fast Reactor systems. (orig.)

  19. Formulation and evaluation of gas release scenarios for the silo in Swedish Final Repository for Radioactive Waste (SFR)

    International Nuclear Information System (INIS)

    Carlsson, J.; Moreno, L.

    1992-01-01

    The Swedish Final Repository for Radioactive Waste (SFR) has been in operation since 1988 and is located in the crystalline rock, 60 m below the Baltic Sea. In the licensing procedure for the SFR the safety assessment has been complemented with a detailed scenario analysis of the performance of the repository. The scenarios include the influence on radionuclide release by gas formation and gas transport processes in the silo. The overall conclusion is that the release of most radionuclides from the silo is only marginally affected by the formation and release of gas, even for scenarios considering unexpected events. The largest effects were found for short-lived radionuclides and radionuclides that have no or low sorption ability. Except for very extreme scenarios for the silo the overall impact from repository on the environment is by far dominated by the release of radionuclides from the rock vaults. 10 refs., 6 figs

  20. Site investigation SFR. Bedrock geology

    International Nuclear Information System (INIS)

    Curtis, Philip; Markstroem, Ingemar; Petersson, Jesper; Triumf, Carl-Axel; Isaksson, Hans; Mattsson, Haakan

    2011-12-01

    SKB is currently carrying out an assessment of the future extension of the final repository for low and middle level radioactive operational waste, SFR. The planned SFR extension lies at a relatively shallow depth (-50 to -200 masl) compared with the planned Forsmark facility for spent nuclear fuel (-400 to -500 masl). The main aim of the multidisciplinary modelling project involving geology, hydrogeology, hydrogeochemistry and rock mechanical modelling is to describe the rock volume for the planned extension of SFR that was presented in /SKB 2008a/. The results of the modelling project in the form of a forthcoming site descriptive model will supply the basis for site-adapted design including engineering characteristics, in addition to a general assessment of the site suitability. The current report presents the results of the geological work with the deterministic rock domain and deformation zone models (version 1.0) and forms a basis for the three other disciplines in the modelling work. The shallow depth of SFR and its proposed extension means that the facility lies partly within the rock volume affected by the effects of stress release processes during loading and unloading cycles, with an associated increased frequency of open sub-horizontal fractures in the near-surface realm (above -150 masl) compared with that observed at greater depths. The main report describes the data input to the modelling work, the applied modelling methodology and the overall results. More detailed descriptions of the individual modelled deformation zones and rock domains are included in the appendices. The geological modelling work during version 1.0 follows SKB's established methodology using the Rock Visualisation System (RVS). The deformation zone model version 1.0 is a further development of the previous version 0.1 /Curtis et al. 2009/. While the main input to deformation zone model version 0.1 was older geological data from the construction of SFR, including drawings of the

  1. Site investigation SFR. Bedrock geology

    Energy Technology Data Exchange (ETDEWEB)

    Curtis, Philip; Markstroem, Ingemar (Golder Associates AB (Sweden)); Petersson, Jesper (Vattenfall Power Consultant AB (Sweden)); Triumf, Carl-Axel; Isaksson, Hans; Mattsson, Haakan (GeoVista AB (Sweden))

    2011-12-15

    SKB is currently carrying out an assessment of the future extension of the final repository for low and middle level radioactive operational waste, SFR. The planned SFR extension lies at a relatively shallow depth (-50 to -200 masl) compared with the planned Forsmark facility for spent nuclear fuel (-400 to -500 masl). The main aim of the multidisciplinary modelling project involving geology, hydrogeology, hydrogeochemistry and rock mechanical modelling is to describe the rock volume for the planned extension of SFR that was presented in /SKB 2008a/. The results of the modelling project in the form of a forthcoming site descriptive model will supply the basis for site-adapted design including engineering characteristics, in addition to a general assessment of the site suitability. The current report presents the results of the geological work with the deterministic rock domain and deformation zone models (version 1.0) and forms a basis for the three other disciplines in the modelling work. The shallow depth of SFR and its proposed extension means that the facility lies partly within the rock volume affected by the effects of stress release processes during loading and unloading cycles, with an associated increased frequency of open sub-horizontal fractures in the near-surface realm (above -150 masl) compared with that observed at greater depths. The main report describes the data input to the modelling work, the applied modelling methodology and the overall results. More detailed descriptions of the individual modelled deformation zones and rock domains are included in the appendices. The geological modelling work during version 1.0 follows SKB's established methodology using the Rock Visualisation System (RVS). The deformation zone model version 1.0 is a further development of the previous version 0.1 /Curtis et al. 2009/. While the main input to deformation zone model version 0.1 was older geological data from the construction of SFR, including drawings of

  2. Project SAFE. Complexing agents in SFR

    Energy Technology Data Exchange (ETDEWEB)

    Fanger, G.; Skagius, K.; Wiborgh, M. [Kemakta Konsult AB, Stockholm (Sweden)

    2001-01-01

    Low- and intermediate level radioactive waste, produced at Swedish nuclear power plants, will be deposited in an underground repository, SFR. Different substances in the waste or in degradation products emanating from the waste, and chemicals added during the building of cementitious barriers in the repository, may exhibit complexing properties. The complexation of radionuclides with such ligands may increase the mobility of the deposited radionuclides as sorption on the cement phases is decreased and solubility increased. This could lead to an increased leaching of the radionuclides from the repository to the geosphere and biosphere. To be able to evaluate the implications for the function and long-term safety of the repository a study has been performed on complexants in SFR. The study is a part of project SAFE (Safety Assessment of Final Repository for operational Radioactive Waste) at the Swedish Nuclear Fuel and Waste Management Co, SKB. Concentrations of complexants were calculated in different waste types in the repository and compared to critical levels above which radionuclide sorption may be affected. The analysis is based on recent research presented in international and national literature sources. The waste in SFR that may act or give rise to substances with complexing properties mainly consists of cellulose materials, including cement additives used in waste conditioning and backfill grout. The radioactive waste also contains chemicals mainly used in decontamination processes at the nuclear power plants, e.g. EDTA, NTA, gluconate, citric acid and oxalic acid. The calculations performed in this report show that the presence of complexants in SFR may lead to a sorption reduction for some radionuclides in certain waste types. This may have to be considered when performing calculations of the radionuclide transport. Concentration calculations of isosaccharinic acid (ISA), using a degradation yield of 0.1 mole/kg cellulose (2%), showed that the limit above

  3. Project SAFE. Complexing agents in SFR

    International Nuclear Information System (INIS)

    Fanger, G.; Skagius, K.; Wiborgh, M.

    2001-01-01

    Low- and intermediate level radioactive waste, produced at Swedish nuclear power plants, will be deposited in an underground repository, SFR. Different substances in the waste or in degradation products emanating from the waste, and chemicals added during the building of cementitious barriers in the repository, may exhibit complexing properties. The complexation of radionuclides with such ligands may increase the mobility of the deposited radionuclides as sorption on the cement phases is decreased and solubility increased. This could lead to an increased leaching of the radionuclides from the repository to the geosphere and biosphere. To be able to evaluate the implications for the function and long-term safety of the repository a study has been performed on complexants in SFR. The study is a part of project SAFE (Safety Assessment of Final Repository for operational Radioactive Waste) at the Swedish Nuclear Fuel and Waste Management Co, SKB. Concentrations of complexants were calculated in different waste types in the repository and compared to critical levels above which radionuclide sorption may be affected. The analysis is based on recent research presented in international and national literature sources. The waste in SFR that may act or give rise to substances with complexing properties mainly consists of cellulose materials, including cement additives used in waste conditioning and backfill grout. The radioactive waste also contains chemicals mainly used in decontamination processes at the nuclear power plants, e.g. EDTA, NTA, gluconate, citric acid and oxalic acid. The calculations performed in this report show that the presence of complexants in SFR may lead to a sorption reduction for some radionuclides in certain waste types. This may have to be considered when performing calculations of the radionuclide transport. Concentration calculations of isosaccharinic acid (ISA), using a degradation yield of 0.1 mole/kg cellulose (2%), showed that the limit above

  4. Optimized design of an ex-vessel cooling thermosyphon for decay heat removal in SFR

    International Nuclear Information System (INIS)

    Choi, Jae Young; Jeong, Yong Hoon; Song, Sub Lee; Chang, Soon Heung

    2017-01-01

    Passive decay heat removal and sodium fire are two major key issues of nuclear safety in sodium-cooled fast reactor (SFR). Several decay heat removal systems (DHR) were suggested for SFR around the world so far. Those DHRS mainly classified into two concepts: Direct reactor cooling system and ex-vessel cooling system. Direct reactor cooling method represented by PDHRS from PGSFR has disadvantages on its additional in-vessel structure and potential sodium fire risk due to the sodium-filled heat exchanger exposed to air. Contrastively, ex-vessel cooling method represented by RVACS from PRISM has low decay heat removal performance, which cannot be applicable to large scale reactors, generally over 1000 MWth. No passive DHRSs which can solve both side of disadvantages has been suggested yet. The goal of this study was to propose ex-vessel cooling system using two-phase closed thermosyphon to compensate the disadvantages of the past DHRSs. Reference reactor was Innovative SFR (iSFR), a pool-type SFR designed by KAIST and featured by extended core lifetime and increased thermal efficiency. Proposed ex-vessel cooling system consisted of 4 trains of thermosyphons and designed to remove 1% of thermal power with 10% of margin. The scopes of this study were design of proposed passive DHRS, validation of system analysis and optimization of system design. Mercury was selected as working fluid to design ex-vessel thermosyphon in consideration of system geometry, operating temperature and required heat flux. SUS 316 with chrome coated liner was selected as case material to resist against high corrosivity of mercury. Thermosyphon evaporator was covered on the surface of reactor vessel as the geometry of hollow shell filled with mercury. Condenser was consisted of finned tube bundles and was located in isolated water pool, the ultimate heat sink. Operation limits and thermal resistance was estimated to guarantee whether the design was adequate. System analysis was conducted by in

  5. SKB's Project SAFE for the SFR 1 Repository. A Review by Consultants to SKI

    International Nuclear Information System (INIS)

    Chapman, N.A.; Maul, P.R.; Robinson, P.C.; Savage, D.

    2002-06-01

    The SFR 1 repository used for final disposal of low- and intermediate level radioactive waste produced by the Swedish nuclear power programme, industry, medicine and research. In 1992 it was granted a full-scale operating permit following additional reporting on long-term safety aspects by SKB, including the first in-depth safety assessment in 1991. It was stipulated as part of the full-scale licence for SFR 1 that a revised safety assessment should be carried out by SKB at least every ten years during the continued operation of the facility. The first 10-year SKB re-evaluation, called 'Project SAFE', was submitted to the regulators in 2001. The review of Project SAFE presented in this report is the culmination of several years' work with SKI including: 1. The extension and application of SKI's 'systems' approach to set up a description of the SFR 1 repository using Process Influence Diagrams (PIDs). 2. Participation in the development of a flexible Performance Assessment (PA) software tool (the AMBER code) that enables time-dependent analyses to be made of system behaviour. 3. Use of the PID database to explore, from first principles, issues that are likely to be important in the safety performance of SFR 1 and thereby to identify topics to be explored by PA modelling. 4. Peer review of the main SKB Project SAFE supporting documentation to evaluate quality, completeness and the implications of the results. 5. An independent PA exercise, using the AMBER code. 6. A review of an English translation of Section 5 of SKB's Project SAFE Final Safety Report. The present report covers only items 3 to 6, and a separate report provides a more detailed description of item 5. As a result of this review, the key issues that the regulatory authorities will need to address when reviewing SKB's safety case for SFR 1 have been identified as: 1. There is no clear statement of SKB's overall safety concept for SFR 1. It is therefore difficult to judge the results of the PA against

  6. Weld Joint Design for SFR Metallic Fuel Element Closures

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jung Won; Kim, Soo Sung; Woo, Yoon Myeng; Kim, Hyung Tae; Kim, Ki Hwan; Yoon, Kyung Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    The sodium-cooled fast reactor (SFR) system is among the six systems selected for Gen-IV promising systems and expected to become available for commercial introduction around 2030. In Korea, the R and D on SFR has been begun since 1997, as one of the national long-term nuclear R and D programs. The international collaborative research is under way on fuel developments within Advanced Fuel Project for Gen-IV SFR with the closed fuel cycle of full actinide recycling, while TRU bearing metallic fuel, U-TRU-Zr alloy fuel, was selected and is being developed. For the fabrication of SFR metallic fuel elements, the endplug welding is a crucial process. The sealing of endplug to cladding tube should be hermetically perfect to prevent a leakage of fission gases and to maintain a good reactor performance. In this study, the joint designs for endplug welding were investigated. For the irradiation test of SFR metallic fuel element, the TIG welding technique was adopted and the welding joint design was developed based on the welding conditions and parameters established. In order to make SFR metallic fuel elements, the weld joint design was developed based on the TIG welding technique.

  7. Site investigation SFR. Water-rock interaction and mixing modelling in the SFR

    Energy Technology Data Exchange (ETDEWEB)

    Gimeno, Maria J.; Auque, Luis F.; Gomez, Javier B.; Acero, Patricia (University of Zaragoza (Spain))

    2011-10-15

    During 2008, the Swedish Nuclear Fuel and Waste Management Company (SKB) initiated an investigation programme for a future expansion of the final repository for low and medium level radioactive operational waste, SFR, located about 150 km north of Stockholm. The purpose of the investigations was to define and characterise a bedrock volume large enough to allow further storage of operational waste from existing Swedish nuclear power plants and future waste from the decommissioning and dismantling of nuclear power plant reactors (SKB 2008). Of several alternatives, a selected location was investigated southwest of the present SFR tunnel system. As part of the SFR Site Descriptive Model, the objective of the hydrogeochemical site description is to describe the chemistry, origin and distribution of groundwaters in the bedrock and the hydrogeochemical processes involved in their evolution. Hydrogeochemical information (salinity distribution, groundwater residence time, palaeohydrogeochemical input, etc.) are also of importance to help constrain the hydrogeological descriptive model. The hydrogeochemical modelling work has been performed in three steps, resulting in three model versions (0.1, 0.2 and 1.0). In versions 0.1 and 0.2, explorative analyses using traditional geochemical approaches (trend plots, x-y scatter plots, 3D visualisations, etc.) were performed to describe the data and to provide an early insight and understanding of the site. The final hydrogeochemical site description version 1.0 (Nilsson et al. 2011) includes data from the previous versions, as well as subsequent complementary data from the SFR extension project, and all these data are further evaluated using additional modelling approaches and techniques. In this context, the present report gives a more detailed analysis of the available data for some hydrogeochemical systems and a detailed description of the results of the geochemical and statistical modelling. One of the main aims is to establish

  8. Site investigation SFR. Water-rock interaction and mixing modelling in the SFR

    International Nuclear Information System (INIS)

    Gimeno, Maria J.; Auque, Luis F.; Gomez, Javier B.; Acero, Patricia

    2011-10-01

    During 2008, the Swedish Nuclear Fuel and Waste Management Company (SKB) initiated an investigation programme for a future expansion of the final repository for low and medium level radioactive operational waste, SFR, located about 150 km north of Stockholm. The purpose of the investigations was to define and characterise a bedrock volume large enough to allow further storage of operational waste from existing Swedish nuclear power plants and future waste from the decommissioning and dismantling of nuclear power plant reactors (SKB 2008). Of several alternatives, a selected location was investigated southwest of the present SFR tunnel system. As part of the SFR Site Descriptive Model, the objective of the hydrogeochemical site description is to describe the chemistry, origin and distribution of groundwaters in the bedrock and the hydrogeochemical processes involved in their evolution. Hydrogeochemical information (salinity distribution, groundwater residence time, palaeohydrogeochemical input, etc.) are also of importance to help constrain the hydrogeological descriptive model. The hydrogeochemical modelling work has been performed in three steps, resulting in three model versions (0.1, 0.2 and 1.0). In versions 0.1 and 0.2, explorative analyses using traditional geochemical approaches (trend plots, x-y scatter plots, 3D visualisations, etc.) were performed to describe the data and to provide an early insight and understanding of the site. The final hydrogeochemical site description version 1.0 (Nilsson et al. 2011) includes data from the previous versions, as well as subsequent complementary data from the SFR extension project, and all these data are further evaluated using additional modelling approaches and techniques. In this context, the present report gives a more detailed analysis of the available data for some hydrogeochemical systems and a detailed description of the results of the geochemical and statistical modelling. One of the main aims is to establish

  9. Radioactive Waste Generation in Pyro-SFR Nuclear Fuel Cycle

    International Nuclear Information System (INIS)

    Gao, Fanxing; Park, Byung Heung; Ko, Won Il

    2011-01-01

    Which nuclear fuel cycle option to deploy is of great importance in the sustainability of nuclear power. SFR fuel cycle employing pyroprocessing (named as Pyro- SFR Cycle) is one promising fuel cycle option in the near future. Radioactive waste generation is a key criterion in nuclear fuel cycle system analysis, which considerably affects the future development of nuclear power. High population with small territory is one special characteristic of ROK, which makes the waste management pretty important. In this study, particularly the amount of waste generation with regard to the promising advanced fuel cycle option was evaluated, because the difficulty of deploying an underground repository for HLW disposal requires a longer time especially in ROK

  10. Sodium-cooled fast reactor (SFR) fuel assembly design with graphite-moderating rods to reduce the sodium void reactivity coefficient

    Energy Technology Data Exchange (ETDEWEB)

    Won, Jong Hyuck; Cho, Nam Zin, E-mail: nzcho@kaist.ac.kr; Park, Hae Min; Jeong, Yong Hoon, E-mail: jeongyh@kaist.ac.kr

    2014-12-15

    Highlights: • The graphite rod-inserted SFR fuel assembly is proposed to achieve low sodium void reactivity. • The neutronics/thermal-hydraulics analyses are performed for the proposed SFR cores. • The sodium void reactivity is improved about 960–1030 pcm compared to reference design. - Abstract: The concept of a graphite-moderating rod-inserted sodium-cooled fast reactor (SFR) fuel assembly is proposed in this study to achieve a low sodium void reactivity coefficient. Using this concept, two types of SFR cores are analyzed; the proposed SFR type 1 core has new SFR fuel assemblies at the inner/mid core regions while the proposed SFR type 2 core has a B{sub 4}C absorber sandwich in the middle of the active core region as well as new SFR fuel assemblies at the inner/mid core regions. For the proposed SFR core designs, neutronics and thermal-hydraulic analyses are performed using the DIF3D, REBUS3, and the MATRA-LMR codes. In the neutronics analysis, the sodium void reactivity coefficient is obtained in various void situations. The two types of proposed core designs reduce the sodium void reactivity coefficient by about 960–1030 pcm compared to the reference design. However, the TRU enrichment for the proposed SFR core designs is increased. In the thermal hydraulic analysis, the temperature distributions are calculated for the two types of proposed core designs and the mass flow rate is optimized to satisfy the design constraints for the highest power generating assembly. The results of this study indicate that the proposed SFR assembly design concept, which adopts graphite-moderating rods which are inserted into the fuel assembly, can feasibly minimize the sodium void reactivity coefficient. Single TRU enrichment and an identical fuel slug diameter throughout the SFR core are also achieved because the radial power peak can be flattened by varying the number of moderating rods in each core region.

  11. Sodium fast reactor gaps analysis of computer codes and models for accident analysis and reactor safety.

    Energy Technology Data Exchange (ETDEWEB)

    Carbajo, Juan (Oak Ridge National Laboratory, Oak Ridge, TN); Jeong, Hae-Yong (Korea Atomic Energy Research Institute, Daejeon, Korea); Wigeland, Roald (Idaho National Laboratory, Idaho Falls, ID); Corradini, Michael (University of Wisconsin, Madison, WI); Schmidt, Rodney Cannon; Thomas, Justin (Argonne National Laboratory, Argonne, IL); Wei, Tom (Argonne National Laboratory, Argonne, IL); Sofu, Tanju (Argonne National Laboratory, Argonne, IL); Ludewig, Hans (Brookhaven National Laboratory, Upton, NY); Tobita, Yoshiharu (Japan Atomic Energy Agency, Ibaraki-ken, Japan); Ohshima, Hiroyuki (Japan Atomic Energy Agency, Ibaraki-ken, Japan); Serre, Frederic (Centre d' %C3%94etudes nucl%C3%94eaires de Cadarache %3CU%2B2013%3E CEA, France)

    2011-06-01

    This report summarizes the results of an expert-opinion elicitation activity designed to qualitatively assess the status and capabilities of currently available computer codes and models for accident analysis and reactor safety calculations of advanced sodium fast reactors, and identify important gaps. The twelve-member panel consisted of representatives from five U.S. National Laboratories (SNL, ANL, INL, ORNL, and BNL), the University of Wisconsin, the KAERI, the JAEA, and the CEA. The major portion of this elicitation activity occurred during a two-day meeting held on Aug. 10-11, 2010 at Argonne National Laboratory. There were two primary objectives of this work: (1) Identify computer codes currently available for SFR accident analysis and reactor safety calculations; and (2) Assess the status and capability of current US computer codes to adequately model the required accident scenarios and associated phenomena, and identify important gaps. During the review, panel members identified over 60 computer codes that are currently available in the international community to perform different aspects of SFR safety analysis for various event scenarios and accident categories. A brief description of each of these codes together with references (when available) is provided. An adaptation of the Predictive Capability Maturity Model (PCMM) for computational modeling and simulation is described for use in this work. The panel's assessment of the available US codes is presented in the form of nine tables, organized into groups of three for each of three risk categories considered: anticipated operational occurrences (AOOs), design basis accidents (DBA), and beyond design basis accidents (BDBA). A set of summary conclusions are drawn from the results obtained. At the highest level, the panel judged that current US code capabilities are adequate for licensing given reasonable margins, but expressed concern that US code development activities had stagnated and that the

  12. Uncertainty Evaluation of Reactivity Coefficients for a large advanced SFR Core Design

    International Nuclear Information System (INIS)

    Khamakhem, Wassim; Rimpault, Gerald

    2008-01-01

    Sodium Cooled Fast Reactors are currently being reshaped in order to meet Generation IV goals on economics, safety and reliability, sustainability and proliferation resistance. Recent studies have led to large SFR cores for a 3600 MWth power plants, cores which exhibit interesting features. The designs have had to balance between competing aspects such as sustainability and safety characteristics. Sustainability in neutronic terms is translated into positive breeding gain and safety into rather low Na void reactivity effects. The studies have been done on two SFR concepts using oxide and carbide fuels. The use of the sensitivity theory in the ERANOS determinist code system has been used. Calculations have been performed with different sodium evaluations: JEF2.2, ERALIB-1 and the most recent JEFF3.1 and ENDF/B-VII in order to make a broad comparison. Values for the Na void reactivity effect exhibit differences as large as 14% when using the different sodium libraries. Uncertainties due to nuclear data on the reactivity coefficients were performed with BOLNA variances-covariances data, the Na Void Effect uncertainties are near to 12% at 1σ. Since, the uncertainties are far beyond the target accuracy for a design achieving high performance, two directions are envisaged: the first one is to perform new differential measurements or in a second attempt use integral experiments to improve effectively the nuclear data set and its uncertainties such as performed in the past with ERALIB1. (authors)

  13. Characterisation of bitumenised waste in SFR 1

    International Nuclear Information System (INIS)

    Pettersson, Michael; Elert, M.

    2001-06-01

    The waste deposited in the Final Repository for Radioactive Operational Waste, SFR, consists in part of waste solidified in bitumen. Bitumen is considered to have favourable chemical and physical properties to act as a fixation material for radioactive waste. However, during interim storage and subsequent disposal bitumen's properties may change. This may influence the stability of the bitumen matrix to retain radionuclides. This report discusses different processes affecting the long-term performance of bitumenised waste, and an evaluation of these properties in waste deposited in SFR 1 is made. The possible effect of a bitumen barrier on the release rate of radionuclides from SFR 1 is assessed. Based on leaching experiments reviewed in this study, it could take some thousand years, possibly more, to release all radionuclides in a 200-litre drum. The results are, however, extrapolated from experiments performed during a short period of time. Long- term deteriorating effects and the effect of a low temperature on the bitumen matrix are not very well documented. The literature focuses principally on bitumenised evaporator concentrate, but the bitumenised waste deposited in SFR 1 consists mainly of ion exchange resins. There are indications that the non-radioactive waste products usually investigated overestimate bitumen's ability to retain waste. Radiolytic effects has been estimated in this work to be negligible for waste categories F.17, F.20 and B.20 deposited in SFR 1, but for categories B.05, B.06 and F.18 the possibility of increased water uptake rate due to radiolysis can not be excluded. A more reasonable assumption is that bitumen will act as an effective barrier for radionuclide release during a time span from some hundreds to thousand of years. Generally, the majority of the inventory of radionuclides in SFR 1 is not solidified in bitumen. By taking the bitumen barrier into account in the modelling of release of radio- nuclides from SFR 1, the total

  14. Characterisation of bitumenised waste in SFR 1

    Energy Technology Data Exchange (ETDEWEB)

    Pettersson, Michael; Elert, M. [Kemakta Konsult AB, Stockholm (Sweden)

    2001-06-01

    The waste deposited in the Final Repository for Radioactive Operational Waste, SFR, consists in part of waste solidified in bitumen. Bitumen is considered to have favourable chemical and physical properties to act as a fixation material for radioactive waste. However, during interim storage and subsequent disposal bitumen's properties may change. This may influence the stability of the bitumen matrix to retain radionuclides. This report discusses different processes affecting the long-term performance of bitumenised waste, and an evaluation of these properties in waste deposited in SFR 1 is made. The possible effect of a bitumen barrier on the release rate of radionuclides from SFR 1 is assessed. Based on leaching experiments reviewed in this study, it could take some thousand years, possibly more, to release all radionuclides in a 200-litre drum. The results are, however, extrapolated from experiments performed during a short period of time. Long- term deteriorating effects and the effect of a low temperature on the bitumen matrix are not very well documented. The literature focuses principally on bitumenised evaporator concentrate, but the bitumenised waste deposited in SFR 1 consists mainly of ion exchange resins. There are indications that the non-radioactive waste products usually investigated overestimate bitumen's ability to retain waste. Radiolytic effects has been estimated in this work to be negligible for waste categories F.17, F.20 and B.20 deposited in SFR 1, but for categories B.05, B.06 and F.18 the possibility of increased water uptake rate due to radiolysis can not be excluded. A more reasonable assumption is that bitumen will act as an effective barrier for radionuclide release during a time span from some hundreds to thousand of years. Generally, the majority of the inventory of radionuclides in SFR 1 is not solidified in bitumen. By taking the bitumen barrier into account in the modelling of release of radio- nuclides from SFR 1, the

  15. Endplug Welding Techniques developed for SFR Metallic Fuel Elements

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jung Won; Kim, Soo Sung; Woo, Yoon Myeng; Kim, Hyung Tae; Lee, Ho Jin; Kim, Ki Hwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    In Korea, the R and D on SFR has been begun since 1997, as one of the national long-term nuclear R and D programs. The international collaborative research is under way on fuel developments within Advanced Fuel Project for Gen-IV SFR with the closed fuel cycle of full actinide recycling, while TRU bearing metallic fuel, U-TRU-Zr alloy fuel, was selected and is being developed. For the fabrication of SFR metallic fuel elements, the endplug welding is a crucial process. The sealing of endplug to cladding tube should be hermetically perfect to prevent a leakage of fission gases and to maintain a good reactor performance. In this study, the welding technique, welding equipment, welding conditions and parameters were developed to make SFR metallic fuel elements. The TIG welding technique was adopted and the welding joint design was developed. And the optimal welding conditions and parameters were also established. In order to make SFR metallic fuel elements, the welding technique, welding equipment, welding conditions and parameters were developed. The TIG welding technique was adopted and the welding joint design was developed. And the optimal welding conditions and parameters were also established.

  16. Endplug Welding Techniques developed for SFR Metallic Fuel Elements

    International Nuclear Information System (INIS)

    Lee, Jung Won; Kim, Soo Sung; Woo, Yoon Myeng; Kim, Hyung Tae; Lee, Ho Jin; Kim, Ki Hwan

    2013-01-01

    In Korea, the R and D on SFR has been begun since 1997, as one of the national long-term nuclear R and D programs. The international collaborative research is under way on fuel developments within Advanced Fuel Project for Gen-IV SFR with the closed fuel cycle of full actinide recycling, while TRU bearing metallic fuel, U-TRU-Zr alloy fuel, was selected and is being developed. For the fabrication of SFR metallic fuel elements, the endplug welding is a crucial process. The sealing of endplug to cladding tube should be hermetically perfect to prevent a leakage of fission gases and to maintain a good reactor performance. In this study, the welding technique, welding equipment, welding conditions and parameters were developed to make SFR metallic fuel elements. The TIG welding technique was adopted and the welding joint design was developed. And the optimal welding conditions and parameters were also established. In order to make SFR metallic fuel elements, the welding technique, welding equipment, welding conditions and parameters were developed. The TIG welding technique was adopted and the welding joint design was developed. And the optimal welding conditions and parameters were also established

  17. Implementation of project Safe in Amber. Verification study for SFR 1 SAR-08

    Energy Technology Data Exchange (ETDEWEB)

    Thomson, Gavin; Herben, Martin; Lloyd, Pam; Rose, Danny; Smith, Chris; Barraclough, Ian (Enviros Consulting Ltd (GB))

    2008-03-15

    This report documents an exercise in which AMBER has been used to represent the models used in Project SAFE, a safety assessment undertaken on SFR 1. (AMBER is a flexible, graphical-user-interface based tool that allows users to build their own dynamic compartmental models to represent the migration, degradation and fate of contaminants in an environmental system. AMBER allows the user to assess routine, accidental and long-term contaminant release.) AMBER has been used to undertake assessment calculations on all of the disposal system, including all disposal tunnels and the Silo, the geosphere and several biosphere modules. The near-field conceptual models were implemented with minimal changes to the approach undertaken previously in Project SAFE. Model complexity varied significantly between individual disposal facilities increasing significantly from the BLA to the BTF and BMA tunnels and Silo. Radionuclide transport through the fractured granite geosphere was approximated using a compartment model approach in AMBER. Several biosphere models were implemented in AMBER including reasonable biosphere development, which considered the evolution of the Forsmark area from coastal to lacustrine to agricultural environments in response to land uplift. Parameters were sampled from distributions and simulations were run for 1,000 realisations. In undertaking the comparison of AMBER with the various codes and calculation tools used in Project SAFE it was necessary to undertake a detailed analysis of the modelling approach previously adopted, with particular focus given to the near-field models. As a result some discrepancies in the implementation of the models and documentation were noted. The exercise demonstrates that AMBER is fully capable of representing the features of the SFR 1 disposal system in a safety assessment suitable for SAR-08

  18. Implementation of project Safe in Amber. Verification study for SFR 1 SAR-08

    International Nuclear Information System (INIS)

    Thomson, Gavin; Herben, Martin; Lloyd, Pam; Rose, Danny; Smith, Chris; Barra clough, Ian

    2008-03-01

    This report documents an exercise in which AMBER has been used to represent the models used in Project SAFE, a safety assessment undertaken on SFR 1. (AMBER is a flexible, graphical-user-interface based tool that allows users to build their own dynamic compartmental models to represent the migration, degradation and fate of contaminants in an environmental system. AMBER allows the user to assess routine, accidental and long-term contaminant release.) AMBER has been used to undertake assessment calculations on all of the disposal system, including all disposal tunnels and the Silo, the geosphere and several biosphere modules. The near-field conceptual models were implemented with minimal changes to the approach undertaken previously in Project SAFE. Model complexity varied significantly between individual disposal facilities increasing significantly from the BLA to the BTF and BMA tunnels and Silo. Radionuclide transport through the fractured granite geosphere was approximated using a compartment model approach in AMBER. Several biosphere models were implemented in AMBER including reasonable biosphere development, which considered the evolution of the Forsmark area from coastal to lacustrine to agricultural environments in response to land uplift. Parameters were sampled from distributions and simulations were run for 1,000 realisations. In undertaking the comparison of AMBER with the various codes and calculation tools used in Project SAFE it was necessary to undertake a detailed analysis of the modelling approach previously adopted, with particular focus given to the near-field models. As a result some discrepancies in the implementation of the models and documentation were noted. The exercise demonstrates that AMBER is fully capable of representing the features of the SFR 1 disposal system in a safety assessment suitable for SAR-08

  19. Design of FCI Experiments to Understand Fuel Out-Pin Phenomena in the SFR

    Energy Technology Data Exchange (ETDEWEB)

    Heo, Hyo; Park, Seong Dae [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of); Jerng, Dong Wook; Bang, In Cheol [Chungang Univ., Seoul (Korea, Republic of)

    2014-05-15

    It is important to guarantee a passive nuclear safety regarding enhanced negative reactivity by fragmenting the molten fuel. In the SFR, it has a strong point that the negative reactivity is immediately introduced when the metal fuel is melted by the UTOP or ULOP accident. These characteristics of the metal fuel can prevent from progressing in severe accidents such as core disruptive accidents (CDA). As key phenomena in the accidents, fuel-coolant interaction (FCI) phenomena have been studied over the last few decades. Especially, several previous researches focused on instability and fragmentation of a core melt jet in water. However, the studies showed too limited phenomena to fully understand. In the domestic SFR technology development, researches for severe accidents tend to lag behind ones of other countries. Or, South Korea has a very basic level of the research such as literature survey. Recently, the SAS4A code, which was developed at Argonne National Laboratory (ANL) for thermal-hydraulic and neutronic analyses of power and flow transients in liquid-metal-cooled nuclear reactors (LMRs), is still under development to consider for a metal fuel. The other countries carried out basic experiments for molten fuel and coolant interactions. However, in a high temperature condition, methods for analysis of structural interaction between molten fuel and fuel cladding are very limited. The ultimate objective of the study is to evaluate the possibility of recriticality accident induced by fuel-coolant interaction in the SFR adopting metal fuel. It is a key point to analyze the molten-fuel behavior based on the experimental results which show fuel-coolant interaction with the simulant materials. It is necessary to establish the test facility, to build database, and to develop physical models to understand the FCI phenomena in the SFR; molten fuel-coolant interaction as soon as the molten fuel is ejected to the sodium coolant channel and molten fuel-coolant interaction

  20. Dose assessments for SFR 1

    International Nuclear Information System (INIS)

    Bergstroem, Ulla; Avila, Rodolfo; Ekstroem, Per-Anders; Cruz, Idalmis de la

    2008-05-01

    Following a review by the Swedish regulatory authorities of the safety analysis of the SFR 1 disposal facility for low and intermediate level waste, SKB has prepared an updated safety analysis, SAR-08. This report presents estimations of annual doses to the most exposed groups from potential radionuclide releases from the SFR 1 repository for a number of calculation cases, selected using a systematic approach for identifying relevant scenarios for the safety analysis. The dose estimates can be used for demonstrating that the long term safety of the repository is in compliance with the regulatory requirements. In particular, the mean values of the annual doses can be used to estimate the expected risks to the most exposed individuals, which can then be compared with the regulatory risk criteria for human health. The conversion from doses to risks is performed in the main report. For one scenario however, where the effects of an earthquake taking place close to the repository are analysed, risk calculations are presented in this report. In addition, prediction of concentrations of radionuclides in environmental media, such as water and soil, are compared with concentration limits suggested by the Erica-project as a base for estimating potential effects on the environment. The assessment of the impact on non-human biota showed that the potential impact is negligible. Committed collective dose for an integration period of 10,000 years for releases occurring during the first thousand years after closure are also calculated. The collective dose commitment was estimated to be 8 manSv. The dose calculations were carried out for a period of 100,000 years, which was sufficient to observe peak doses in all scenarios considered. Releases to the landscape and to a well were considered. The peaks of the mean annual doses from releases to the landscape are associated with C-14 releases to a future lake around year 5,000 AD. In the case of releases to a well, the peak annual doses

  1. Dose assessments for SFR 1

    Energy Technology Data Exchange (ETDEWEB)

    Bergstroem, Ulla (Swedish Nuclear Fuel and Waste Management Co., Stockholm (Sweden)); Avila, Rodolfo; Ekstroem, Per-Anders; Cruz, Idalmis de la (Facilia AB, Bromma (Sweden))

    2008-06-15

    Following a review by the Swedish regulatory authorities of the safety analysis of the SFR 1 disposal facility for low and intermediate level waste, SKB has prepared an updated safety analysis, SAR-08. This report presents estimations of annual doses to the most exposed groups from potential radionuclide releases from the SFR 1 repository for a number of calculation cases, selected using a systematic approach for identifying relevant scenarios for the safety analysis. The dose estimates can be used for demonstrating that the long term safety of the repository is in compliance with the regulatory requirements. In particular, the mean values of the annual doses can be used to estimate the expected risks to the most exposed individuals, which can then be compared with the regulatory risk criteria for human health. The conversion from doses to risks is performed in the main report. For one scenario however, where the effects of an earthquake taking place close to the repository are analysed, risk calculations are presented in this report. In addition, prediction of concentrations of radionuclides in environmental media, such as water and soil, are compared with concentration limits suggested by the Erica-project as a base for estimating potential effects on the environment. The assessment of the impact on non-human biota showed that the potential impact is negligible. Committed collective dose for an integration period of 10,000 years for releases occurring during the first thousand years after closure are also calculated. The collective dose commitment was estimated to be 8 manSv. The dose calculations were carried out for a period of 100,000 years, which was sufficient to observe peak doses in all scenarios considered. Releases to the landscape and to a well were considered. The peaks of the mean annual doses from releases to the landscape are associated with C-14 releases to a future lake around year 5,000 AD. In the case of releases to a well, the peak annual doses

  2. TRIUMF - The Swedish data base system for radioactive waste in SFR

    International Nuclear Information System (INIS)

    Skogsberg, Marie; Andersson, Per-Anders

    2006-01-01

    All short lived LLW/ILW from the operation and maintenance of all Swedish Nuclear Power Plants are disposed in SFR, the Swedish final repository for radioactive operational waste. It is important to save all the information about radioactive waste that is needed now and in the future. To be secure that, we have developed a database system in Sweden called Triumf, consisting information about all the waste packages that are disposed in SFR. The waste producers register data concerning individual waste package during production. Before transport to SFR a data file with all information about the individual waste packages is transferred to Triumf. When transferred, the data are checked against accepted limitations before the waste can be loaded on the ship for transport to SFR. After disposal at SFR the deposition location in the repository is added to the database for each waste package. (author)

  3. Improvement of Steam Generator Reliability for GEN-IV SFR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seong O; Kim Se Yun; Kim, Seok Hoon; Eoh, Jae Hyuk; Lee, Hyeong Yeon; Choi, Byung Seon

    2005-11-15

    The R and D items performed in this study were selected from the R and D task of ' Reliability improvement of Steam Generator' of GEN-IV SFR Component Design and BOP. Since this project deals with one of the most important issues for a GEN-IV SFR system, it needs to enhance the domestic technical backgrounds associated with the corresponding R and D items even for a very short period by 2005. This study provides the R and D results for i) Development of assessment methodology for dissimilar metal weld and ii) Development of multi-dimensional simulation methodology for a SWR event in a SFR steam generator.

  4. Improvement of Steam Generator Reliability for GEN-IV SFR

    International Nuclear Information System (INIS)

    Kim, Seong O; Kim Se Yun; Kim, Seok Hoon; Eoh, Jae Hyuk; Lee, Hyeong Yeon; Choi, Byung Seon

    2005-11-01

    The R and D items performed in this study were selected from the R and D task of ' Reliability improvement of Steam Generator' of GEN-IV SFR Component Design and BOP. Since this project deals with one of the most important issues for a GEN-IV SFR system, it needs to enhance the domestic technical backgrounds associated with the corresponding R and D items even for a very short period by 2005. This study provides the R and D results for i) Development of assessment methodology for dissimilar metal weld and ii) Development of multi-dimensional simulation methodology for a SWR event in a SFR steam generator

  5. Multi-physics and multi-objective design of heterogeneous SFR core: development of an optimization method under uncertainty

    International Nuclear Information System (INIS)

    Ammar, Karim

    2014-01-01

    Since Phenix shutting down in 2010, CEA does not have Sodium Fast Reactor (SFR) in operating condition. According to global energetic challenge and fast reactor abilities, CEA launched a program of industrial demonstrator called ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration), a reactor with electric power capacity equal to 600 MW. Objective of the prototype is, in first to be a response to environmental constraints, in second demonstrates the industrial viability of SFR reactor. The goal is to have a safety level at least equal to 3. generation reactors. ASTRID design integrates Fukushima feedback; Waste reprocessing (with minor actinide transmutation) and it linked industry. Installation safety is the priority. In all cases, no radionuclide should be released into environment. To achieve this objective, it is imperative to predict the impact of uncertainty sources on reactor behaviour. In this context, this thesis aims to develop new optimization methods for SFR cores. The goal is to improve the robustness and reliability of reactors in response to existing uncertainties. We will use ASTRID core as reference to estimate interest of new methods and tools developed. The impact of multi-Physics uncertainties in the calculation of the core performance and the use of optimization methods introduce new problems: How to optimize 'complex' cores (i.e. associated with design spaces of high dimensions with more than 20 variable parameters), taking into account the uncertainties? What is uncertainties behaviour for optimization core compare to reference core? Taking into account uncertainties, optimization core are they still competitive? Optimizations improvements are higher than uncertainty margins? The thesis helps to develop and implement methods necessary to take into account uncertainties in the new generation of simulation tools. Statistical methods to ensure consistency of complex multi-Physics simulation results are also

  6. Innovative power conversion system for the French SFR prototype, ASTRID

    International Nuclear Information System (INIS)

    Cachon, L.; Biscarrat, C.; Morin, F.; Haubensack, D.; Rigal, E.; Moro, I.; Baque, F.; Madeleine, S.; Rodriguez, G.; Laffont, G.

    2012-01-01

    In the framework of the French Act of 28 June 2006 about nuclear materials and waste management, the prototype ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration), foreseen in operation by the 20's, will have to demonstrate not only the minor actinide transmutation capability, but also the progress made in Sodium Fast Reactor (SFR) technology on an industrial scale, by qualifying innovative options. Some of these options still require improvements, especially in the field of operability and safety. In fact, one of the main issues with the standard steam/water Power Conversion System (PCS) of SFR is the fast and energetic chemical reaction between water and sodium, which could occur in steam generators in case of tube failure. To manage the sodium/water reaction, one way consists in minimizing the impact of such event: hence studies are carried out on steam generator design, improvement of the physical knowledge of this phenomenon, development of numerical simulation to predict the reaction onset and consequences, and associated detection improvement. On the other hand, the other way consists in eliminating sodium/water reaction. In this frame, the CEA contribution to the feasibility evaluation of an alternative innovative PCS (replacing steam/water by 180 bar pressurised nitrogen) is focused on the following main topics: - The parametric study leading to nitrogen selection: the thermodynamic cycle efficiency optimisation on Brayton cycles is performed with several gases at different pressures. - The design of innovative compact heat exchangers for the gas loop: here the key points are the nuclear codification associated with inspection capability, the innovative welding process and the thermal-hydraulic and thermal-mechanic optimisations. After a general introduction of the ASTRID project, this paper presents in detail these different feasibility studies being led on the innovative gas PCS for an SFR. (authors)

  7. Prediction of the Sodium Void Reactivity in the Metal-fueled SFR Using the ENDF/B-VII.0 Library

    Energy Technology Data Exchange (ETDEWEB)

    Yun, Sunghwan; Lim, Jae-Yong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The SVR (Sodium Void Reactivity) is one of the most important parameters in SFR (Sodium-cooled Fast Reactor) safety analysis. In this paper, to estimate the error of the SVR in metal-fueled SFR, three physics experiments named as BFS-75-1, BFS-109-2A, and BFS-84-1 were examined using recent cross-section library, ENDF/B-VII.0 and the MCNP code. In the MCNP6 calculation, two million histories/generation with 50 inactive/300 active generations are used with the continuous-energy ENDF/B-VII.0 library. We expect that accuracy of total cross-section of the sodium may play a dominant role in errors of SVRs at core peripheral and sodium plenum regions, whereas accuracy of capture cross-section of the sodium may play a dominant role for the results in errors of SVRs at core central region. In addition, capture cross-sections of the sodium in the ENDF/B-VII.0, the JEFF-3.2, and the JENDL-4.0 libraries show significant differences between each other, while total cross-sections of sodium in three libraries show good agreement.

  8. DNA homologous recombination factor SFR1 physically and functionally interacts with estrogen receptor alpha.

    Directory of Open Access Journals (Sweden)

    Yuxin Feng

    Full Text Available Estrogen receptor alpha (ERα, a ligand-dependent transcription factor, mediates the expression of its target genes by interacting with corepressors and coactivators. Since the first cloning of SRC1, more than 280 nuclear receptor cofactors have been identified, which orchestrate target gene transcription. Aberrant activity of ER or its accessory proteins results in a number of diseases including breast cancer. Here we identified SFR1, a protein involved in DNA homologous recombination, as a novel binding partner of ERα. Initially isolated in a yeast two-hybrid screen, the interaction of SFR1 and ERα was confirmed in vivo by immunoprecipitation and mammalian one-hybrid assays. SFR1 co-localized with ERα in the nucleus, potentiated ER's ligand-dependent and ligand-independent transcriptional activity, and occupied the ER binding sites of its target gene promoters. Knockdown of SFR1 diminished ER's transcriptional activity. Manipulating SFR1 expression by knockdown and overexpression revealed a role for SFR1 in ER-dependent and -independent cancer cell proliferation. SFR1 differs from SRC1 by the lack of an intrinsic activation function. Taken together, we propose that SFR1 is a novel transcriptional modulator for ERα and a potential target in breast cancer therapy.

  9. Modification of the SAS4A Safety Analysis Code for Integration with the ADAPT Discrete Dynamic Event Tree Framework.

    Energy Technology Data Exchange (ETDEWEB)

    Jankovsky, Zachary Kyle [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Denman, Matthew R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2017-05-01

    It is difficult to assess the consequences of a transient in a sodium-cooled fast reactor (SFR) using traditional probabilistic risk assessment (PRA) methods, as numerous safety-related sys- tems have passive characteristics. Often there is significant dependence on the value of con- tinuous stochastic parameters rather than binary success/failure determinations. One form of dynamic PRA uses a system simulator to represent the progression of a transient, tracking events through time in a discrete dynamic event tree (DDET). In order to function in a DDET environment, a simulator must have characteristics that make it amenable to changing physical parameters midway through the analysis. The SAS4A SFR system analysis code did not have these characteristics as received. This report describes the code modifications made to allow dynamic operation as well as the linking to a Sandia DDET driver code. A test case is briefly described to demonstrate the utility of the changes.

  10. Evolution of near-field physico-chemical characteristics of the SFR repository

    Energy Technology Data Exchange (ETDEWEB)

    Savage, D [Quintessa Ltd., Nottingham (United Kingdom); Stenhouse, M [Monitor Scientific LLC, Denver, CO (United States); Benbow, S [Quintessa Ltd., Henley-on-Thames (United Kingdom)

    2000-08-01

    The evaluation of the post-closure performance of the SFR repository needs to consider time dependent evolution of the repository environment. Time-dependent reaction of near-field barriers (cement, steel, bentonite) with saturating groundwater will lead to the development of hyper alkaline repository pore fluids, chemically reducing conditions, and ultimately, the generation of gas through anaerobic corrosion of metals. Cement and concrete will act as chemical conditioning agents to minimise metal corrosion and ultimately, maximise radioelement sorption. The chemical and physical evolution of cement and concrete through reaction with ambient groundwater will thus affect sorption processes through changes in pH, complexing ligands, and solid surface properties. It is desirable that these changes be incorporated into the safety assessment. The sorption behaviour of radionuclides in cementitious systems has been reviewed in detail. The available evidence from experimental work carried out on the influence of organic materials on the sorption behaviour of radionuclides, indicates that most organic degradation products will not affect sorption significantly at the concentrations expected in a cementitious repository. The notable exception to this conclusion involves the degradation products of cellulose and, in particular, polycarboxylic acids represented by iso-saccharinic acid (ISA). Results using ISA indicate a significant reduction in sorption of Pu, by several orders of magnitude, for an ISA concentration of about 10{sup -3} M. More recent data indicate that the negative effect is not as great, though still significant. Therefore, some scoping calculations are advisable to determine how realistic an ISA concentration of about 10{sup -3} M would be for the SFR repository and to estimate concentrations of other relevant organic compounds, in particular EDTA, for comparison. Scoping calculations relevant to the longevity of hyper alkaline pore fluid conditions at SFR

  11. ASN’s actions in GEN IV reactors and Sodium Fast Reactors (SFR)

    International Nuclear Information System (INIS)

    Belot, Clotilde

    2013-01-01

    The ASN is involved in 3 actions concerning GEN IV: • Overview of nuclear reactor GEN IV systems; • Specific analysis about transmutation; • Prototype reactor ASTRID (SFR). Furthermore theses actions are in the beginning (no conclusions or results available)

  12. Integrated CFD investigation of heat transfer enhancement using multi-tray core catcher in SFR

    International Nuclear Information System (INIS)

    Rakhi; Sharma, Anil Kumar; Velusamy, K.

    2017-01-01

    Highlights: • Heat transfer enhancement using multi-tray core catcher for SFR is investigated. • The capability of a single core collector tray is estimated. • Double and triple collector trays with innovative designs is discussed. • Provision of openings in the trays contributed to enhanced natural circulation. - Abstract: To render future SFR more robust and safe, certain BDBE have been considered in the recent years. A Core Disruptive Accident leading to a whole core meltdown scenario has gained the interest of researchers. Various design concepts and safety measures have been suggested and incorporated in design to address such a low probability scenario. A core catcher concept, in particular, has proved to be inevitable as an in-vessel core retention device in SFR for safe retention of core debris arising out after the severe accident. This study aims to analyse the cooling capability of the innovative design concept of core catcher to remove decay heat of degraded core after the accident. First, the capability of single collection tray is established and then the study is extended to two and three collection trays with different design concepts. Transient forms of governing equations of mass, momentum and energy conservations along with k-ε turbulence model are solved by finite volume based CFD solver. Boussinesq approximation is invoked to model buoyancy in sodium. The study shows that a single collection tray is capable of removing up to 20 MW decay heat load in a typical 500 MWe pool type SFR. Further, studies are carried out to improve the natural circulation of sodium around the source, in the lower plenum and to distribute core debris of the whole core to multiple collection trays. It is found that the double and triple collection trays can accommodate decay loads up to 29 MW. Provision of openings in the collection trays has proved to be effective in improving the heat transfer and sodium flow as well as in distributing the core debris to the

  13. Korean SFR development program and technical activities for improving economical competitiveness

    International Nuclear Information System (INIS)

    Yoo, Jaewoon

    2013-01-01

    Future Plan: • Construction cost evaluation of PGSFR and commercial SFR; – Component based capital cost evaluation of PGSFR is undergoing and will be completed by the first half of 2014; – Component cost is only based on the experience from that of LWR; • Cost Benefit Analysis of Future Nuclear Energy Mix; – With revised National Energy Plan (as of 2013); – Near-term: Benefit from LWR spent fuel recycling: - In Korean law, Share of Expense for spent fuel disposal is reserved as 0.4M$ per a LWR spent fuel assembly (as of 2003); – Long-term: Competitive power plant to LWR with self sustainable feature; • Revision of commercial SFR conceptual design; – Less constraint in material (fuel, cladding) irradiation experience; – More innovative features as long-term goal

  14. Groundwater chemical changes at SFR in Forsmark

    Energy Technology Data Exchange (ETDEWEB)

    Laaksoharju, Marcus [GeoPoint AB, Sollentuna (Sweden); Gurban, Ioana [3DTerra (Sweden)

    2003-01-01

    The examination of the groundwater sampled at the SFR tunnel system indicated that the groundwater consist mainly of a Na-Cl to Na-Ca-Cl type of water. Most of the samples fall within the Cl range of 2500-5500 mg/l having a neutral pH (6.6-7.7 units). The water is reducing and despite the fact that the tunnel acts like a hydraulic sink constantly withdrawing water out from the rock into the tunnel the groundwater changes are moderate with time. Most of the sampling points in the SFR tunnel system are located under the Sea and M3 calculations indicated that most of the sampling points have a change of water types from an older marine water type affected by glacial melt water to an more modern marine water type such as Baltic Sea water which has been modified by possibly microbial sulphate reduction and ion exchange. Mass balance calculations indicated that the waters seem to be in equilibrium with the fracture filling mineral such as calcite. The quality of the aluminium data made the modelling with the major rock forming aluminium silicates such as feldspars and clay minerals uncertain and was therefore not reported. The conclusion is that the groundwater evolution and patterns at SFR are a result of many factors such as: 1. the changes in hydrogeology related to glaciation/deglaciation and land uplift, 2. repeated Sea/lake water regressions/transgressions 3. the closeness to Baltic Sea resulting in relative small hydrogeological driving forces which could preserve old water types from being flushed out, 4. organic or inorganic alteration of the groundwater caused by microbial processes or in situ water/rock interactions 5. tunnel construction which changed the flow system The modelled present-day groundwater conditions of the SFR site consist of a mixture in varying degrees of different water types. The data indicate that all the groundwater at SFR is strongly affected by Sea water of different origin and ages. The meteoric (0- 1000 B.P) portion is located close

  15. Evolution of near-field physico-chemical characteristics of the SFR repository

    International Nuclear Information System (INIS)

    Savage, D.; Stenhouse, M.; Benbow, S.

    2000-08-01

    The evaluation of the post-closure performance of the SFR repository needs to consider time dependent evolution of the repository environment. Time-dependent reaction of near-field barriers (cement, steel, bentonite) with saturating groundwater will lead to the development of hyper alkaline repository pore fluids, chemically reducing conditions, and ultimately, the generation of gas through anaerobic corrosion of metals. Cement and concrete will act as chemical conditioning agents to minimise metal corrosion and ultimately, maximise radioelement sorption. The chemical and physical evolution of cement and concrete through reaction with ambient groundwater will thus affect sorption processes through changes in pH, complexing ligands, and solid surface properties. It is desirable that these changes be incorporated into the safety assessment. The sorption behaviour of radionuclides in cementitious systems has been reviewed in detail. The available evidence from experimental work carried out on the influence of organic materials on the sorption behaviour of radionuclides, indicates that most organic degradation products will not affect sorption significantly at the concentrations expected in a cementitious repository. The notable exception to this conclusion involves the degradation products of cellulose and, in particular, polycarboxylic acids represented by iso-saccharinic acid (ISA). Results using ISA indicate a significant reduction in sorption of Pu, by several orders of magnitude, for an ISA concentration of about 10 -3 M. More recent data indicate that the negative effect is not as great, though still significant. Therefore, some scoping calculations are advisable to determine how realistic an ISA concentration of about 10 -3 M would be for the SFR repository and to estimate concentrations of other relevant organic compounds, in particular EDTA, for comparison. Scoping calculations relevant to the longevity of hyper alkaline pore fluid conditions at SFR have been

  16. The terrestrial biosphere in the SFR region

    Energy Technology Data Exchange (ETDEWEB)

    Jerling, L; Isaeus, M [Stockholm Univ. (Sweden). Dept. of Botany; Lanneck, J [Stockholm Univ. (Sweden). Dept. of Physical Geography; Lindborg, T; Schueldt, R [Danish Nature Council, Copenhagen (Denmark)

    2001-03-01

    This report is a part of the SKB project 'SAFE' (Safety Assessment of the Final Repository of Radioactive Operational Waste). The aim of project SAFE is to update the previous safety analysis of SFR-1.SFR-1 is a facility for disposal of low and intermediate level radioactive waste, which is situated in bedrock beneath the Baltic Sea, one km off the coast near the Forsmark nuclear power plant in Northern Uppland. A part of the SAFE-analysis aims at analysing the transport of radionuclides in the ecosystems.To do so one has to build a model that includes a large amount of information concerning the biosphere.The first step is to collect and compile descriptions of the biosphere.This report is a first attempt to characterise the terrestrial environment of the SFR area of Forsmark. In the first part of the report the terrestrial environment, land class distribution and production of the area is described. The primary production in different terrestrial ecosystems is estimated for a model area in the Forsmark region. The estimations are based on the actual land class distribution and the values for the total primary production (d.w. above ground biomass)and the amount carbon produced, presented as g/m{sup 2} for each land class respectively. An important aspect of the biosphere is the vegetation and its development. The future development of vegetation is of interest since production,decomposition and thus storage of organic material, vary strongly among vegetation types and this has strong implications for the transport of radionuclides.Therefore an attempt to describe the development of terrestrial vegetation has been made in the second part. Any prediction of future vegetation is based on knowledge of the past together with premises for the future development.The predictions made, thus, becomes marred with errors enforced by the assumptions and incomplete information of the past. The assumptions made for the predictions in this report are crude and results in a

  17. The terrestrial biosphere in the SFR region

    International Nuclear Information System (INIS)

    Jerling, L.; Isaeus, M.

    2001-03-01

    This report is a part of the SKB project 'SAFE' (Safety Assessment of the Final Repository of Radioactive Operational Waste). The aim of project SAFE is to update the previous safety analysis of SFR-1.SFR-1 is a facility for disposal of low and intermediate level radioactive waste, which is situated in bedrock beneath the Baltic Sea, one km off the coast near the Forsmark nuclear power plant in Northern Uppland. A part of the SAFE-analysis aims at analysing the transport of radionuclides in the ecosystems.To do so one has to build a model that includes a large amount of information concerning the biosphere.The first step is to collect and compile descriptions of the biosphere.This report is a first attempt to characterise the terrestrial environment of the SFR area of Forsmark. In the first part of the report the terrestrial environment, land class distribution and production of the area is described. The primary production in different terrestrial ecosystems is estimated for a model area in the Forsmark region. The estimations are based on the actual land class distribution and the values for the total primary production (d.w. above ground biomass)and the amount carbon produced, presented as g/m 2 for each land class respectively. An important aspect of the biosphere is the vegetation and its development. The future development of vegetation is of interest since production,decomposition and thus storage of organic material, vary strongly among vegetation types and this has strong implications for the transport of radionuclides.Therefore an attempt to describe the development of terrestrial vegetation has been made in the second part. Any prediction of future vegetation is based on knowledge of the past together with premises for the future development.The predictions made, thus, becomes marred with errors enforced by the assumptions and incomplete information of the past. The assumptions made for the predictions in this report are crude and results in a coarse

  18. The terrestrial biosphere in the SFR region

    Energy Technology Data Exchange (ETDEWEB)

    Jerling, L.; Isaeus, M. [Stockholm Univ. (Sweden). Dept. of Botany; Lanneck, J. [Stockholm Univ. (Sweden). Dept. of Physical Geography; Lindborg, T.; Schueldt, R. [Danish Nature Council, Copenhagen (Denmark)

    2001-03-01

    This report is a part of the SKB project 'SAFE' (Safety Assessment of the Final Repository of Radioactive Operational Waste). The aim of project SAFE is to update the previous safety analysis of SFR-1.SFR-1 is a facility for disposal of low and intermediate level radioactive waste, which is situated in bedrock beneath the Baltic Sea, one km off the coast near the Forsmark nuclear power plant in Northern Uppland. A part of the SAFE-analysis aims at analysing the transport of radionuclides in the ecosystems.To do so one has to build a model that includes a large amount of information concerning the biosphere.The first step is to collect and compile descriptions of the biosphere.This report is a first attempt to characterise the terrestrial environment of the SFR area of Forsmark. In the first part of the report the terrestrial environment, land class distribution and production of the area is described. The primary production in different terrestrial ecosystems is estimated for a model area in the Forsmark region. The estimations are based on the actual land class distribution and the values for the total primary production (d.w. above ground biomass)and the amount carbon produced, presented as g/m{sup 2} for each land class respectively. An important aspect of the biosphere is the vegetation and its development. The future development of vegetation is of interest since production,decomposition and thus storage of organic material, vary strongly among vegetation types and this has strong implications for the transport of radionuclides.Therefore an attempt to describe the development of terrestrial vegetation has been made in the second part. Any prediction of future vegetation is based on knowledge of the past together with premises for the future development.The predictions made, thus, becomes marred with errors enforced by the assumptions and incomplete information of the past. The assumptions made for the predictions in this report are crude and results

  19. A comparison of in-vessel behaviors between SFR and PWR under severe accident

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sanggil; Cho, Cheon Hwey [ACT Co., Daejeon (Korea, Republic of); Kim, Sang Ji [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    This paper aims to provide an easy guide for experts who know well the severe accident phenomenology of Pressurized Water Reactor (PWR) by comparing both reactor design concepts and in vessel behaviors under a postulated severe accident condition. This study only provides a preliminary qualitative comparison based on available literature. The PWR and SFR in-vessel design concepts and their effects under a postulate severe accident are investigated in this paper. Although this work is a preliminary study to compare the in-vessel behaviors for both PWR and SFR, it seems that there is no possibility to lead a significant core damage in the metal fuel SFR concept. In the oxide fuel SFR, there might be a chance to progress to the severe accident initiators such as the energetic reaction, flow blockage and so on.

  20. Boron-bearing Influences of 9Cr-0.5Mo-2W-V-Nb Ferritic/Martensitic Steels for a SFR Fuel Cladding

    International Nuclear Information System (INIS)

    Baek, Jong-Hyuk; Han, Chang-Hee; Kim, Woo-Gon; Kim, Sung-Ho; Lee, Chan-Bock

    2008-01-01

    Currently the principal materials in a SFR (sodium-cooled fast reactor) of Gen-IV nuclear system are considering stainless steels (e.g. austenitic steels and ferritic/martensitic steels) for pressure boundary and structural applications in the primary circuit (cladding, duct, cold and hot leg piping, and pressure vessel). There are sound technical justifications for these material selections, and the adoption of these stainless steels for a wide range of nuclear and non-nuclear applications has generated much industrial technology and experience. However, there are strong incentives to develop advanced materials, especially cladding, for the Gen-IV SFR. The Gen-IV SFR is to have a considerable increase in safety and be economically competitive when compared with the conventional water reactors. To accomplish these objectives, the development of the fuel cladding material should be set forth as a premise because its integrity is directly related to those of the reactor system as well as the fuel in the Gen-IV SFR. Since last year, a R and D program was launched to develop the improved ferritic/martensitic steel for the Gen-IV SFR fuel cladding. Categories of materials considered in the program included 8 - 12% Cr ferritic/ martensitic steels. A strong recommendation was made for the development of a high strength steel equivalent to or superior to ASTM Gr.92 steel to offset the difficulties encountered with commercial available steels of the 8 - 12% Cr group. That is, since fuel cladding in the Gen-IV SFR would operate under higher temperatures than 600 .deg. C, contacting with liquid sodium, and be irradiated by neutrons to as high as 200dpa, the cladding should thus sustain both superior irradiation and temperature stabilities during an operational life. The newly developed advanced steel should overcome the severe drawback; mechanical properties, especially creep, are deteriorated at a higher temperature over 600 .deg. C. In this study, as one of the composition

  1. International review on safety requirements for the prototype fast breeder reactor “Monju”

    International Nuclear Information System (INIS)

    2016-01-01

    In response to the lessons learned from the serious nuclear accidents at the TEPCO's Fukushima Daiichi Nuclear Power Stations, an advisory committee, which was set up by the Japan Atomic Energy Agency, issued the report “Safety Requirements Expected to the Prototype Fast Breeder Reactor Monju” taking into account the SFR specific safety characteristics in July 2014. The report was reviewed by the leading international experts on SFR safety from five countries and one international organization in order to obtain independent and objective evaluation. The international review comments on each subsection were collected and compiled, and then a summary of results was derived through the discussion at the review meeting and individual feedbacks. As a result the basic concept for prevention of severe accidents and mitigation of their consequences of Monju is appropriate in consideration of SFR specific safety characteristics, and is in accordance with international common understanding. (author)

  2. Challenges in mechanical modeling of SFR fuel rod transient behavior

    Energy Technology Data Exchange (ETDEWEB)

    Feria, F.; Herranz, L. E.

    2013-07-01

    Modeling of SFR fuel rod mechanical behavior under transient conditions entails the development of a creep law to predict cladding viscoplastic strain. In this regard, this work is focused on defining a proper clad creep law structure as the basis to set a suitable model under SFR off-normal conditions as transient overpower and loss of fluid. To do so, a review of in-codes clad creep models has been done by using SAS-SFR, SCANAIR and ASTEC. The proposed creep model has been structured in two parts: viscoplastic behaviour before the failure (primary and secondary creep) and the failure due to viscoplastic collapse (tertiary creep). In order to model the first part, Norton creep law has been proposed as a conservative option. An irradiation hardening factor should be included for best estimate calculations. The recommendation for the second part is to apply a failure criterion based on strain limit or rupture time, which allows achieving conservative results.

  3. Analysis of power ramp rate and minimum power controllability of the MMS model for a plant dynamics analysis of a Prototype SFR

    International Nuclear Information System (INIS)

    Kim, Eui Kwang; Kim, Dehee; Joo, Hyungkook; Lee, Taeho

    2014-01-01

    A full plant dynamic model was developed for a prototype SFR using the Modular Modeling System (MMS). It includes the modeling of various subsystems such as the neutronics, primary and intermediate sodium systems of the NSSS, steam and water systems of the BOP, BOP controls, and the supervisory plant controls. The NSSS model is subdivided into component models, such as a Core, IHXs, Pumps, SGs, and the rest of the NSSS loop model. The BOP model is subdivided into a steam subsystem, feedwater subsystem, and preheater subsystem. Plant transient tests were performed to study the operational considerations. It includes varying the power ramp rate and studying the controllability at minimum power. Plant transient tests were performed to study operational considerations by using the MMS model for a prototype SFR. It includes varying the power ramp rate, studying the controllability at the minimum power set point. At a power ramp rate of higher than 2%, the steam temperature has a large deviation from the target. As the power set point decreases, the PHTS hot leg temperature and steam temperature tend to have higher deviations. After further refinement of the MMS model, it can be useful for developing the plant operation logics of the prototype SFR

  4. Construction of in-situ creep strain test facility for the SFR fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sang Gyu; Heo, Hyeong Min; Kim, Jun Hwan; Kim, Sung Ho [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    In this study, in-situ laser inspection creep test machine was developed for the measuring the creep strain of SFR fuel cladding materials. Ferritic-martensitic steels are being considered as an attractive candidate material for a fuel cladding of a SFR due to their low expansion coefficients, high thermal conductivities and excellent irradiation resistances to a void swelling. HT9 steel (12CrMoVW) is initially developed as a material for power plants in Europe in the 1960. This steel has experienced to expose up to 200dpa in FFTE and EBR-II. Ferritic-Martensitic steel's maximum creep strength in existence is 180Mpa for 106 hour 600 .deg., but HT9 steel is 60Mpa. Because SFR is difficult to secure in developing and applying materials, HT9 steel has accumulated validated data and is suitable for SFR component. And also, because of its superior dimensional stability against fast neutron irradiation, Ferritic-martensitic steel of 9Cr and 12Cr steels, such as HT9 and FC92(12Cr-2W) are preferable to utilize in the fuel cladding of an SFR in KAERI. The pressurized thermal creep test of HT9 and FC92 claddings are being conducted in KAERI, but the change of creep strain in cladding is not easy to measure during the creep test due to its pressurized and closed conditions. In this paper, in-situ laser inspection pressurized creep test machine developed for SFR fuel cladding specimens is described. Moreover, the creep strain rate of HT9 at 650 .deg. C was examined from the in-situ laser inspection pressurized creep test machine.

  5. International review on safety requirements for the prototype fast breeder reactor “Monju” (Translated document)

    International Nuclear Information System (INIS)

    2016-02-01

    In response to the lessons learned from the serious nuclear accidents at the TEPCO's Fukushima Daiichi Nuclear Power Stations, an advisory committee, which was set up by the Japan Atomic Energy Agency, issued the report “Safety Requirements Expected to the Prototype Fast Breeder Reactor Monju” taking into account the SFR specific safety characteristics in July 2014. The report was reviewed by the leading international experts on SFR safety from five countries and one international organization in order to obtain independent and objective evaluation. The international review comments on each subsection were collected and compiled, and then a summary of results was derived through the discussion at the review meeting and individual feedbacks. As a result the basic concept for prevention of severe accidents and mitigation of their consequences of Monju is appropriate in consideration of SFR specific safety characteristics, and is in accordance with international common understanding. (author)

  6. Measurement and flow visualization research of thermal hydraulic characteristics for the SFR reactor Vessel

    International Nuclear Information System (INIS)

    Cha, J. E.; Kim, S. O.; Choi, H. L.; Kim, H. B.; Kim, H. W.; Lee, S. H.

    2012-01-01

    In this report, the thermal hydraulic and flow visualization experiment was described for the KALIMER-600 water-scaled model. In order to investigate a thermal hydraulic characteristics for the SFR KALIMER-600, which has been conceptually designed in the KAERI, a water-scaled 1/10 reactor vessel model was designed and prepared through the scaling analysis during three-years research. In this research, SFR Photos system, which has inherently very complicated the internal structures, was fabricated with a transparent vessel. It was shown that a serious of thermal hydraulic test was conducted within a short period if modeled with water than sodium. Natural circulation test was successfully performed with the modeled heater assembly and heat exchanger system coupled with cooling system. The water-scaled RSV experimental facility made in this research could be used to study the USA development for the future SFR system and utilized to analyze the flow characteristics before changing a main internal part of Photos system. It could also be used to test a pool-inspection study and a sensor selection study before large scale sodium experiment. The PCV system prepared in this research could be utilized to test other TSH experiment and temperature field measurement

  7. Development of Melting Crucible Materials of Metallic Fuel Slug for SFR

    International Nuclear Information System (INIS)

    Kim, K. H.; Lee, C. T.; Oh, S. J.; Kim, S. K.; Lee, C. B.; Ko, Y. M.; Woo, W. M.

    2010-01-01

    The fabrication process of metallic fuel for SFR(sodium fast reactor) of Generation-IV candidate reactors is composed of the fabrication of fuel pin, fuel rod, and fuel assembly. The key technology of the fabrication process for SFR can be referred to the fabrication technology of fuel pin. As SFR fuel contains MA(minor actinide) elements proceeding the recycling of actinide elements, it is so important to extinguish MA during irradiation in SFR, included in nuclear fuel through collection of volatile MA elements during fabrication of fuel pin. Hence, it is an imminent circumstance to develop the fabrication process of fuel pin. This report is an state-of art report related to the characteristics of irradiation performance for U-Zr-Pu metallic fuel, and the apparatus and the technology of conventional injection casting process. In addition, to overcome the drawbacks of the conventional injection casting and the U-Zr-Pu fuel, new fabrication technologies such as the gravity casting process, the casting of fuel pin to metal-barrier mold, the fabrication of particulate metallic fuel utilizing centrifugal atomization is surveyed and summarized. The development of new U-10Mo-X metallic fuel as nuclear fuel having a single phase in the temperature range between 550 and 950 .deg. C, reducing the re-distribution of the fuel elements and improving the compatibility between fuel and cladding, is also surveyed and summarized

  8. Evaluation of core modeling effect on transients for multi-flow zone design of SFR

    International Nuclear Information System (INIS)

    Shin, Andong; Choi, Yong Won

    2016-01-01

    SFR core is composed of different types of assemblies including fuel driver, reflector, blanket, control, safety drivers and other drivers. Modeling of different types of assemblies is inevitable in general. But modeling of core flow zones of with different channels needs a lot of effort and could be a challenge for system code modeling due to its limitation on the number of modeling components. In this study, core modeling effect on SFR transient was investigated with flow-zone model and averaged inner core channel model to improve modeling efficiency and validation of simplified core model for EBR-II loss of flow transient case with the modified TRACE code for SFRs. Core modeling effect on the loss flow transient was analyzed with flow-zoned channel model, single averaged inner core model and highest flow channel with averaged inner core channel model for EBR-II SHRT-17 test core. Case study showed that estimations of transient pump and channel flow as well as channel outlet temperatures were similar for all cases macroscopically. Comparing the result of the base case (flow-zone channel inner core model) and the case 2 (highest flow channel considered averaged inner core channel model), flow and channel outlet temperature response were closer than the case1 (single averaged inner core model)

  9. Evaluation of core modeling effect on transients for multi-flow zone design of SFR

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Andong; Choi, Yong Won [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2016-10-15

    SFR core is composed of different types of assemblies including fuel driver, reflector, blanket, control, safety drivers and other drivers. Modeling of different types of assemblies is inevitable in general. But modeling of core flow zones of with different channels needs a lot of effort and could be a challenge for system code modeling due to its limitation on the number of modeling components. In this study, core modeling effect on SFR transient was investigated with flow-zone model and averaged inner core channel model to improve modeling efficiency and validation of simplified core model for EBR-II loss of flow transient case with the modified TRACE code for SFRs. Core modeling effect on the loss flow transient was analyzed with flow-zoned channel model, single averaged inner core model and highest flow channel with averaged inner core channel model for EBR-II SHRT-17 test core. Case study showed that estimations of transient pump and channel flow as well as channel outlet temperatures were similar for all cases macroscopically. Comparing the result of the base case (flow-zone channel inner core model) and the case 2 (highest flow channel considered averaged inner core channel model), flow and channel outlet temperature response were closer than the case1 (single averaged inner core model)

  10. The evolution of the dust temperatures of galaxies in the SFR-M∗ plane up to z ∼ 2

    Science.gov (United States)

    Magnelli, B.; Lutz, D.; Saintonge, A.; Berta, S.; Santini, P.; Symeonidis, M.; Altieri, B.; Andreani, P.; Aussel, H.; Béthermin, M.; Bock, J.; Bongiovanni, A.; Cepa, J.; Cimatti, A.; Conley, A.; Daddi, E.; Elbaz, D.; Förster Schreiber, N. M.; Genzel, R.; Ivison, R. J.; Le Floc'h, E.; Magdis, G.; Maiolino, R.; Nordon, R.; Oliver, S. J.; Page, M.; Pérez García, A.; Poglitsch, A.; Popesso, P.; Pozzi, F.; Riguccini, L.; Rodighiero, G.; Rosario, D.; Roseboom, I.; Sanchez-Portal, M.; Scott, D.; Sturm, E.; Tacconi, L. J.; Valtchanov, I.; Wang, L.; Wuyts, S.

    2014-01-01

    We study the evolution of the dust temperature of galaxies in the SFR- M∗ plane up to z ~ 2 using far-infrared and submillimetre observations from the Herschel Space Observatory taken as part of the PACS Evolutionary Probe (PEP) and Herschel Multi-tiered Extragalactic Survey (HerMES) guaranteed time key programmes. Starting from a sample of galaxies with reliable star-formation rates (SFRs), stellar masses (M∗) and redshift estimates, we grid the SFR- M∗parameter space in several redshift ranges and estimate the mean dust temperature (Tdust) of each SFR-M∗ - z bin. Dust temperatures are inferred using the stacked far-infrared flux densities (100-500 μm) of our SFR-M∗ - z bins. At all redshifts, the dust temperature of galaxies smoothly increases with rest-frame infrared luminosities (LIR), specific SFRs (SSFR; i.e., SFR/M∗), and distances with respect to the main sequence (MS) of the SFR- M∗ plane (i.e., Δlog (SSFR)MS = log [SSFR(galaxy)/SSFRMS(M∗,z)]). The Tdust - SSFR and Tdust - Δlog (SSFR)MS correlations are statistically much more significant than the Tdust - LIR one. While the slopes of these three correlations are redshift-independent, their normalisations evolve smoothly from z = 0 and z ~ 2. We convert these results into a recipe to derive Tdust from SFR, M∗ and z, valid out to z ~ 2 and for the stellar mass and SFR range covered by our stacking analysis. The existence of a strong Tdust - Δlog (SSFR)MS correlation provides us with several pieces of information on the dust and gas content of galaxies. Firstly, the slope of the Tdust - Δlog (SSFR)MS correlation can be explained by the increase in the star-formation efficiency (SFE; SFR/Mgas) with Δlog (SSFR)MS as found locally by molecular gas studies. Secondly, at fixed Δlog (SSFR)MS, the constant dust temperature observed in galaxies probing wide ranges in SFR and M∗ can be explained by an increase or decrease in the number of star-forming regions with comparable SFE enclosed in

  11. Minutes of the kick-off Meeting of the WPRS / EGRPANS / Sodium Fast Reactor Task Force (SFR)

    International Nuclear Information System (INIS)

    Buiron, Laurent; Stauff, Nicolas; Varaine, Frederic; Blanchet, D.; Stauff, N.; Ivanov, Evgeny; Michel-Sendis, Franco; ); Mikityuk, Konstantin; Pelloni, Sandro; Ponomarev, Alexander; Kim, Taek K.; Taiwo, Temitope; Kereszturi, Andras; Van den Eynde, Gert; Kotiluoto, Petri; Juutilainen, Pauli; Lepp Anen, Jaakko

    2011-01-01

    The kick-off meeting of the Sodium-cooled fast reactor Task Force (SFR) was an informal one where discussion among participants was encouraged. F. Varaine reminded the participants of the objectives and milestones of the proposed benchmark. A discussion followed on how to define transients in a safety approach. Speaking for ANL, T.A Taiwo anticipated the need to include experts from ANL safety division in the thermal-hydraulics/transient part of the benchmark and observed the initial timeline proposed was too tight to meet. It was agreed the timeline of the benchmark would be modified to a more realistic schedule, keeping the neutronic part as the first stage to finish before the end of the year. It was decided that CEA would provide a spreadsheet template to fill out by benchmark participants to facilitate compilation of results. During the presentations, L. Buiron presented the benchmark in detail and introduced the bibliographic study he has been compiling with the aim to develop a state of the art of feedback calculations. He reported this work is ongoing and could become a contribution for a report of the Expert Group. E. Ivanov presented a proposal for complementary calculations to be included in the benchmark. He reported these calculations would allow for interpretation of the discrepancies that may be observed between the different results obtained by the participants, and commented such an interpretation is essential in any safety analysis presented before decision-makers. A discussion followed on the necessity to include the additional calculations proposed by IRSN at such an early state of the benchmark. The proposal was met with interest, and it was decided that it would be discussed again at the next meeting where first neutronic results could be shown and the importance of these calculations demonstrated. E. Ivanov agreed to provide more detailed specifications of the tallies requested by the next meeting. It was decided that these complementary

  12. Comparison of lead and sodium-cooled reactors - Safety, fuel cycle performance and some economical aspects

    Energy Technology Data Exchange (ETDEWEB)

    Carlsson, Johan; Tucek, Kamil; Wider, Hartmut [Joint Research Centre, EC-JRC, Westerduinweg 3, P.O. Box 2, NL-0 1755 ZG Petten (Netherlands)

    2006-07-01

    This paper compares the Lead-cooled Fast Reactor (LFR) and the Sodium-cooled Fast Reactor (SFR) regarding different aspects of the coolant, safety and economics. A brief review of design and safety experience of an SFR (BN-600) and some safety philosophy of the most developed LFR (BREST) are presented as well. The pros and cons of the lead and the sodium coolants are discussed. This paper presents results concerning the coolant temperature evolution during three accident scenarios, i.e. Loss-Of- Flow (LOF), Loss-Of-Heat-Sink (LOHS), and Total-Loss-Of-Power (TLOP). It also studies possible moderators, like BeO and hydrides, for the core designs to have negative reactivity feedbacks and favorable reactivity swings. LFR seems to be able to accommodate more minor actinides than SFR at comparable coolant and Doppler feedbacks. We show that LFR can be designed both to breed and burn transuranics from LWRs. The hydrides lead to the most favorable reactivity feedbacks, but the poorest reactivity swing. It is shown that the LFR can handle the LOF transient better than the SFR. This is due to the much lower pressure drop in the LFR core. The coolant outlet temperatures stabilize at 2050 K and 940 K for SFR and LFR, respectively when no feedbacks are considered. Investigations also concern the SFR's performance when the pitch-to-diameter was increased from 1.2 to 1.4. For the LOHS and TLOP accidents their temperature evolutions are milder for the LFR since lead has a 50% larger volumetric heat capacity. For the TLOP the core outlet temperature of the LFR peaks at 1080 K after 2 days. Regarding economics it appears easier to avoid an intermediate cycle in an LFR than an SFR. (authors)

  13. Comparison of lead and sodium-cooled reactors - Safety, fuel cycle performance and some economical aspects

    International Nuclear Information System (INIS)

    Carlsson, Johan; Tucek, Kamil; Wider, Hartmut

    2006-01-01

    This paper compares the Lead-cooled Fast Reactor (LFR) and the Sodium-cooled Fast Reactor (SFR) regarding different aspects of the coolant, safety and economics. A brief review of design and safety experience of an SFR (BN-600) and some safety philosophy of the most developed LFR (BREST) are presented as well. The pros and cons of the lead and the sodium coolants are discussed. This paper presents results concerning the coolant temperature evolution during three accident scenarios, i.e. Loss-Of- Flow (LOF), Loss-Of-Heat-Sink (LOHS), and Total-Loss-Of-Power (TLOP). It also studies possible moderators, like BeO and hydrides, for the core designs to have negative reactivity feedbacks and favorable reactivity swings. LFR seems to be able to accommodate more minor actinides than SFR at comparable coolant and Doppler feedbacks. We show that LFR can be designed both to breed and burn transuranics from LWRs. The hydrides lead to the most favorable reactivity feedbacks, but the poorest reactivity swing. It is shown that the LFR can handle the LOF transient better than the SFR. This is due to the much lower pressure drop in the LFR core. The coolant outlet temperatures stabilize at 2050 K and 940 K for SFR and LFR, respectively when no feedbacks are considered. Investigations also concern the SFR's performance when the pitch-to-diameter was increased from 1.2 to 1.4. For the LOHS and TLOP accidents their temperature evolutions are milder for the LFR since lead has a 50% larger volumetric heat capacity. For the TLOP the core outlet temperature of the LFR peaks at 1080 K after 2 days. Regarding economics it appears easier to avoid an intermediate cycle in an LFR than an SFR. (authors)

  14. Development of Ultrasonic Visual Inspection Program for In-Vessel Structures of SFR

    International Nuclear Information System (INIS)

    Joo, Y. S.; Park, C. G.; Lee, J. H.

    2009-02-01

    As the liquid sodium of a sodium-cooled fast reactor (SFR) is opaque to light, a conventional visual inspection is unavailable for the evaluation of the in-vessel structures under a sodium level. ASME Section XI Division 3 provides rules and guidelines for an in-service inspection (ISI) and testing of the components of SFR. For the ISI of in-vessel structures, the ASME code specifies visual examinations. An ultrasonic wave should be applied for an under-sodium visual inspection of the in-vessel structures. The plate-type waveguide sensor has been developed and the feasibility of the waveguide sensor technique has been successfully demonstrated for an ultrasonic visual inspection of the in-vessel structures of SFR. In this study, the C-scan image mapping program (Under-Sodium MultiView) is developed to apply this waveguide sensor technology to an under-sodium visual inspection of in-vessel structures in SFR by using a LabVIEW graphical programming language. The Under-Sodium MultiVIEW program has the functions of a double rotating scanner motion control, a high power pulser receiver control, a image mapping and a signal processing. The performance of Under-Sodium MultiVIEW program was verified by a C-scanning test

  15. Application of the SPH method in nodal diffusion analyses of SFR cores

    Energy Technology Data Exchange (ETDEWEB)

    Nikitin, Evgeny; Fridman, Emil [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany). Div. Reactor Safety; Mikityuk, K. [Paul Scherrer Institut, Villigen (Switzerland)

    2016-07-01

    The current study investigated the potential of the SPH method, applied to correct the few-group XS produced by Serpent, to further improve the accuracy of the nodal diffusion solutions. The procedure for the generation of SPH-corrected few-group XS is presented in the paper. The performance of the SPH method was tested on a large oxide SFR core from the OECD/NEA SFR benchmark. The reference SFR core was modeled with the DYN3D and PARCS nodal diffusion codes using the SPH-corrected few-group XS generated by Serpent. The nodal diffusion results obtained with and without SPH correction were compared to the reference full-core Serpent MC solution. It was demonstrated that the application of the SPH method improves the accuracy of the nodal diffusion solutions, particularly for the rodded core state.

  16. Operation and Performance of the Supercritical Fluids Reactor (SFR)

    National Research Council Canada - National Science Library

    Hanush, R

    1996-01-01

    The Supercritical Fluids Reactor (SFR) at Sandia National Laboratories, CA has been developed to examine and solve engineering, process, and fundamental chemistry issues regarding the development of supercritical water oxidation (SCWO...

  17. Safety Design Criteria and Approaches to Safety Substantiation of the BN-1200

    International Nuclear Information System (INIS)

    Ashurko, I.

    2013-01-01

    Russian experience in SFR area: Activities on development of safety design criteria for SFRs of the 4th generation is carried out within the GIF framework. Although this reactor technology is considered as innovative that is relevant to the 4th generation, however, it has already a certain history. In this relation, it seems to be useful to analyze the corresponding experience that is available in various countries. 4 SFRs have been successfully operated in the USSR and in the Russian Federation: • Experimental reactor BR-5/10; • Research reactor BOR-60; • Prototype BN-350 power reactor; • Commercial BN-600 power unit at the Beloyarsk NPP. Thus, Russia gained a considerable experience of design, construction and operation of SFRs. In particular, a certain experience has been acquired on safety substantiation of reactors of this type and their licensing. Now BOR-60 and BN-600 continue their operation, BN-800 power unit is under construction, development of the commercial BN-1200 power unit, that is considered as the 4th generation reactor, has been started. Due to limited number of operating SFRs in the world, successful Russian experience in this area should be taken into account for further development and improvement of SFR SDC developed by the GIF Task Force. In particular, participation of SFR designers in this activities would be fruitful and useful

  18. Generation-IV nuclear reactors, SFR concept

    International Nuclear Information System (INIS)

    Dufour, P.

    2010-01-01

    In this presentation author deals with development of sodium-cooled fast reactors and lead-cooled fast reactors. He concluded that: - SFR is a proved concept that has never achieved industrial deployment; - The GEN IV objectives need to reconsider the design of both the core and the reactor design : innovations are being analysed; Future design will benefit from considerable feedback of design, licensing, construction and operation of PX, SPX, etc.

  19. On-going activities in the European JASMIN project for the development and validation of ASTEC-Na SFR safety simulation code - 15072

    International Nuclear Information System (INIS)

    Girault, N.; Cloarec, L.; Herranz, L.; Bandini, G.; Perez-Martin, S.; Ammirabile, L.

    2015-01-01

    The 4-year JASMIN collaborative project (Joint Advanced Severe accidents Modelling and Integration for Na-cooled fast reactors), started in Dec.2011 in the frame of the 7. Framework Programme of the European Commission. It aims at developing a new European simulation code, ASTEC-Na, dealing with the primary phase of SFR core disruptive accidents. The development of a new code, based on a robust advanced simulation tool and able to encompass the in-vessel and in-containment phenomena occurring during a severe accident is indeed of utmost interest for advanced and innovative future SFRs for which an enhanced safety level will be required. This code, based on the ASTEC European code system developed by IRSN and GRS for severe accidents in water-cooled reactors, is progressively integrating and capitalizing the state-of-the-art knowledge of SFR accidents through physical model improvement or development of new ones. New models are assessed on in-pile (CABRI, SCARABEE etc...) and out-of pile experiments conducted during the 70's-80's and code-o-code benchmarking with current accident simulation tools for SFRs is also conducted. During the 2 and a half first years of the project, model specifications and developments were conducted and the validation test matrix was built. The first version of ASTEC-Na available in early 2014 already includes a thermal-hydraulics module able to simulate single and two-phase sodium flow conditions, a zero point neutronic model with simple definition of channel and axial dependences of reactivity feedbacks and models derived from SCANAIR IRSN code for simulating fuel pin thermo-mechanical behaviour and fission gas release/retention. Meanwhile, models have been developed in the source term area for in-containment particle generation and particle chemical transformation, but their implementation is still to be done. As a first validation step, the ASTEC-Na calculations were satisfactorily compared to thermal-hydraulics experimental

  20. Site investigation SFR. Reprocessing of reflection seismic profiles 5b and 8, Forsmark

    Energy Technology Data Exchange (ETDEWEB)

    Juhlin, Christopher; Zhang, Fengjiao (Uppsala Univ., Dept. of Earth Sciences (Sweden))

    2010-12-15

    Reflection seismic profiles 5b and 8 in the northern Forsmark area have been reprocessed with the aim of improving the images in the uppermost 500 metres in the SFR area. The main conclusion is that a new reflection (B10) has been identified that may extend below the SFR site. This reflection was not clearly observed in the previous processing. The reflection strikes approximately N25E and dips at about 35 degrees to the southeast. This orientation is similar to the set B group identified earlier /Juhlin and Palm 2005/. Note that the dip of the reflection is uncertain. On shot gathers it appears to dip at a slightly shallower angle while on the stacked sections it appears to dip at a greater angle. This discrepancy is probably due to the crooked nature of the profiles. However, reflections are clearly observed in shot gathers and its presence below SFR is highly probable. Two new reflections were also identified further north along profile 5b (A11 and A12). These dip to the south-southeast, but would be found at a depth of 1-2 km below SFR if they extend to below the site. There are also signs of a 3rd reflection with similar orientation to the set A group identified earlier, A13, but its existence is very speculative. This reflector would intersect the surface within the SFR area. South of the Singoe deformation zone on profile 5b, another new reflection has been found, N1. The orientation of this reflection is speculative since it is not clearly seen on profile 8. It has been modelled as dipping to the north at about 35 degrees and projects to the surface south of the main SFR area. In addition, the orientation of reflection B7 has been revised as has the lateral extent of A1. Most importantly, A1 is now interpreted not to extend to the surface and not cross the Singoe deformation zone

  1. Hydrogeological flux scenarios at Forsmark. Generic numerical flow simulations and compilation of climatic information for use in the safety analysis SFR1 SAR-08

    International Nuclear Information System (INIS)

    Vidstrand, Patrik; Naeslund, Jens-Ove; Hartikainen, Juha; Svensson, Urban

    2007-11-01

    In the earlier modelling for SFR-SAFE it was concluded that the groundwater flow would increase with time along with the shoreline displacement. Even though the numerical results are different the same conclusion is drawn after this study. General conclusions from the present study are that: The upper boundary conditions have a significant impact on the groundwater flow in the geosphere. The characteristic of the surface in regards of being a recharge or discharge area affects the results. In general, a discharge area will experience an increase in groundwater flow under changed conditions. The presence of caging fracture zones affects the results, and, for the tested un-frozen SFR situation, the resulting effect is an increase in groundwater flow. Specific conclusions regarding the relative change of groundwater flow due to different surface conditions are that: The permafrost scenarios, along with the development from sporadic permafrost to continuous permafrost, yield increased groundwater flows in unfrozen parts of the domain. The increase is one order of magnitude or less. In the permafrost, the flow is negligible. The ice sheet scenarios yield situations with significantly increased groundwater flow. The results indicate an increase by two to three orders of magnitude. These increased values, however, apply only for short duration intervals. It is possible that such intervals may be only a couple of years. In the selected climate Base variant, repeating the conditions for the last glacial cycle, permafrost conditions occur after 8,000 years. In the climate variant affected by increased greenhouse warming, permafrost conditions do not occur until after more than 50,000 years. In the chosen climate variants, ice sheets reach the Forsmark area and cause significantly increased groundwater flow, after ∼60,000 years or more

  2. Hydrogeological flux scenarios at Forsmark. Generic numerical flow simulations and compilation of climatic information for use in the safety analysis SFR1 SAR-08

    Energy Technology Data Exchange (ETDEWEB)

    Vidstrand, Patrik (Bergab, Goeteborg (SE)); Naeslund, Jens-Ove (Swedish Nuclear Fuel and Waste Management Co., Stockholm (SE)); Hartikainen, Juha (Helsinki Univ. of Technology, Helsinki (FI)); Svensson, Urban (CFE AB, Karlskrona (SE))

    2007-11-15

    In the earlier modelling for SFR-SAFE it was concluded that the groundwater flow would increase with time along with the shoreline displacement. Even though the numerical results are different the same conclusion is drawn after this study. General conclusions from the present study are that: The upper boundary conditions have a significant impact on the groundwater flow in the geosphere. The characteristic of the surface in regards of being a recharge or discharge area affects the results. In general, a discharge area will experience an increase in groundwater flow under changed conditions. The presence of caging fracture zones affects the results, and, for the tested un-frozen SFR situation, the resulting effect is an increase in groundwater flow. Specific conclusions regarding the relative change of groundwater flow due to different surface conditions are that: The permafrost scenarios, along with the development from sporadic permafrost to continuous permafrost, yield increased groundwater flows in unfrozen parts of the domain. The increase is one order of magnitude or less. In the permafrost, the flow is negligible. The ice sheet scenarios yield situations with significantly increased groundwater flow. The results indicate an increase by two to three orders of magnitude. These increased values, however, apply only for short duration intervals. It is possible that such intervals may be only a couple of years. In the selected climate Base variant, repeating the conditions for the last glacial cycle, permafrost conditions occur after 8,000 years. In the climate variant affected by increased greenhouse warming, permafrost conditions do not occur until after more than 50,000 years. In the chosen climate variants, ice sheets reach the Forsmark area and cause significantly increased groundwater flow, after approx60,000 years or more

  3. Synthesis of results obtained on sodium components and technology through the Generation IV International Forum SFR Component Design and Balance-of-Plant Project

    International Nuclear Information System (INIS)

    Sienicki, J.J.; Rodriguez, G.; Kisohara, N.; Kim, J. B.; Gerber, A.; Ashurko, Y.; Toyama, S.

    2013-01-01

    Status: The viability of designing SFR components and BOP has been demonstrated with design, construction and operation of previous sodium-cooled reactors. The main objective of this R&D project is related to system performance, or by development on the use of AECS in the BOP that could allow further cost improvements. Objective: To conduct collaborative research and development of components and BOP for the SFR System. The Project has to satisfy the GIF’s criteria of safety, economy, sustainability, proliferation resistance and physical protection. Activities within this Project are addressing experimental and analytical evaluation of advanced ISI&R, LBB assessment, development of AECS with Brayton cycles, advanced SG technologies. Project activities will be based in part on the extensive historical R&D experience with component design and balance of plant for sodium-cooled fast reactors

  4. Development of SFR Research and Integration Management System (S-RIMS)

    International Nuclear Information System (INIS)

    Cho, Chung Ho; Chang, Jin Wook; Kim, Young Gyun; Kim, Yeong Il

    2011-01-01

    Up to the present, the management of research and development (R and D) for a sodium cooled fast reactor (SFR) could be individually performed on each project without an organic relationship. However, a more systemic and effective integrated management of a project is required because the research and development environment is currently changing. Thus, we developed a Research and Integration Management System for SFR (S-RIMS) based on the enterprise project management (EPM) solution. The functional goals of the S-RIMS are as follows: 1. Provide data that show the progress and status of a project 2. Manage the design process and R and D products 3. Share the consistent design data between sub-projects

  5. The SFR-M∗ main sequence archetypal star-formation history and analytical models

    Science.gov (United States)

    Ciesla, L.; Elbaz, D.; Fensch, J.

    2017-12-01

    The star-formation history (SFH) of galaxies is a key assumption to derive their physical properties and can lead to strong biases. In this work, we derive the SFH of main sequence (MS) galaxies and show how the peak SFH of a galaxy depends on its seed mass at, for example, z = 5. This seed mass reflects the galaxy's underlying dark matter (DM) halo environment. We show that, following the MS, galaxies undergo a drastic slow down of their stellar mass growth after reaching the peak of their SFH. According to abundance matching, these masses correspond to hot and massive DM halos which state could result in less efficient gas inflows on the galaxies and thus could be the origin of limited stellar mass growth. As a result, we show that galaxies, still on the MS, can enter the passive region of the UVJ diagram while still forming stars. The best fit to the MS SFH is provided by a right skew peak function for which we provide parameters depending on the seed mass of the galaxy. The ability of the classical analytical SFHs to retrieve the star-formation rate (SFR) of galaxies from spectral energy distribution (SED) fitting is studied. Due to mathematical limitations, the exponentially declining and delayed SFH struggle to model high SFR, which starts to be problematic at z > 2. The exponentially rising and log-normal SFHs exhibit the opposite behavior with the ability to reach very high SFR, and thus model starburst galaxies, but they are not able to model low values such as those expected at low redshift for massive galaxies. By simulating galaxies SED from the MS SFH, we show that these four analytical forms recover the SFR of MS galaxies with an error dependent on the model and the redshift. They are, however, sensitive enough to probe small variations of SFR within the MS, with an error ranging from 5 to 40% depending on the SFH assumption and redshift; but all the four fail to recover the SFR of rapidly quenched galaxies. However, these SFHs lead to an artificial

  6. Site description of the SFR area at Forsmark at completion of the site investigation phase. SDM-PSU Forsmark

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-05-15

    The site descriptive model (SDM) presented in this report is an integrated model for bedrock geology, rock mechanics, bedrock hydrogeology and bedrock hydrogeochemistry of the site investigated in the SFR extension project (PSU). A description of the surface system is also included in the report. However, the surface system is not integrated with the other disciplines as new data regarding the surface system will not be available until after the completion of SDM-PSU. It is noted that SDM-PSU does not include all disciplines handled in SDM-Site Forsmark (SKB 2008b), the focus is to produce a site description that meets the needs of the SFR extension project. The overall objective of the SFR extension project is to have the application for the extension ready by 2013. This report presents an integrated site model incorporating the historic data acquired from the investigations for and construction of the existing SFR facility (1980-1986), as well as from the recent investigations for the planned extension of SFR (2008-2009). It also provides a summary of the abundant underlying data and the discipline-specific models that support the integrated site model. The description relies heavily on background reports concerning detailed data analyses and modelling in the different disciplines. It is noteworthy that the investigations conducted during the SFR extension project were guided by the choice of site prior to the investigations, which was based on the experience gained during the construction of the existing SFR facility.

  7. Site description of the SFR area at Forsmark at completion of the site investigation phase. SDM-PSU Forsmark

    International Nuclear Information System (INIS)

    2013-05-01

    The site descriptive model (SDM) presented in this report is an integrated model for bedrock geology, rock mechanics, bedrock hydrogeology and bedrock hydrogeochemistry of the site investigated in the SFR extension project (PSU). A description of the surface system is also included in the report. However, the surface system is not integrated with the other disciplines as new data regarding the surface system will not be available until after the completion of SDM-PSU. It is noted that SDM-PSU does not include all disciplines handled in SDM-Site Forsmark (SKB 2008b), the focus is to produce a site description that meets the needs of the SFR extension project. The overall objective of the SFR extension project is to have the application for the extension ready by 2013. This report presents an integrated site model incorporating the historic data acquired from the investigations for and construction of the existing SFR facility (1980-1986), as well as from the recent investigations for the planned extension of SFR (2008-2009). It also provides a summary of the abundant underlying data and the discipline-specific models that support the integrated site model. The description relies heavily on background reports concerning detailed data analyses and modelling in the different disciplines. It is noteworthy that the investigations conducted during the SFR extension project were guided by the choice of site prior to the investigations, which was based on the experience gained during the construction of the existing SFR facility

  8. Review of C-14 inventory for the SFR facility

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Graham; Merino, Joan; Kerrigan, Emma

    2002-08-01

    The Swedish Radiation Protection Authority (SSI) is currently reviewing SKB's continuing assessment for disposal of radioactive waste to the SFR facility at Forsmark. Among the wastes disposed are reactor operating wastes. Among the relevant radionuclides is C-14, which is relatively difficult to measure and to control because of its mobility. This report documents a review of the C-14 inventory material submitted by SKB for the SFR-facility, to determine its validity and comment on the appropriate assumptions for C-14 content of wastes due to be disposed of to the SFR. The review is based on information provided by SSI as well as other relevant international experience. Conclusions are drawn upon: the chemical form of the C-14 in the waste from BWRs and PWRs; the production rate of C-14 in BWRs and PWRs and quantification of the source term in the IEX waste; the distribution of the C-14 in the IEX waste from BWR between the resins used for treatment of the primary cooling water and the resins used for treatment of the condensate water; quantification of the uncertainties. A suggestion is made that the C-14 inventory could be better developed based upon a mass balance assessment of all the C-14 produced in reactors, and its ultimate fate in effluent and solid wastes, taking account of the reactor specific operational factors identified in the review as relevant to C-14 inventory assessment.

  9. Review of C-14 inventory for the SFR facility

    International Nuclear Information System (INIS)

    Smith, Graham; Merino, Joan; Kerrigan, Emma

    2002-08-01

    The Swedish Radiation Protection Authority (SSI) is currently reviewing SKB's continuing assessment for disposal of radioactive waste to the SFR facility at Forsmark. Among the wastes disposed are reactor operating wastes. Among the relevant radionuclides is C-14, which is relatively difficult to measure and to control because of its mobility. This report documents a review of the C-14 inventory material submitted by SKB for the SFR-facility, to determine its validity and comment on the appropriate assumptions for C-14 content of wastes due to be disposed of to the SFR. The review is based on information provided by SSI as well as other relevant international experience. Conclusions are drawn upon: the chemical form of the C-14 in the waste from BWRs and PWRs; the production rate of C-14 in BWRs and PWRs and quantification of the source term in the IEX waste; the distribution of the C-14 in the IEX waste from BWR between the resins used for treatment of the primary cooling water and the resins used for treatment of the condensate water; quantification of the uncertainties. A suggestion is made that the C-14 inventory could be better developed based upon a mass balance assessment of all the C-14 produced in reactors, and its ultimate fate in effluent and solid wastes, taking account of the reactor specific operational factors identified in the review as relevant to C-14 inventory assessment

  10. An Application Example Analysis of Quality Assurance Program for STELLA(Sodium Integral Effect Test Loop for Safety Simulation and Assessment) Project

    International Nuclear Information System (INIS)

    Jung, Minhwan; Gam, Dayoung; Eoh, Jae-Hyuk; Jeong, Ji-Young

    2015-01-01

    KAERI has been conducting various basic R and D activities in the field of nuclear technology. In addition, KAERI is now participating in the Generation IV International Forum (GIF), preparing for the development of key technologies for Generation IV nuclear energy system, including Sodium cooled Fast Reactor (SFR) development. All of the key technologies for SFR development need an appropriate level of QA activities to achieve the GIF safety and performance objectives. Therefore, QA activities have been conducted as an essential part of the national SFR project. As a result, QAM (Quality Assurance Manual) and QAP (Quality Assurance Procedures) have been developed for the SFR project, which are based on ASME NQA-1, KEPIC QAP and the GIF Quality Management System Guidelines. In this work, the introduction background and application examples of the QA program for the STELLA project were investigated. Application of the QA for the STELLA project has great significance because the QA has been mainly applied for the nuclear power plant area in operation, which helps ensure the reliability of the test data and completeness of the research performance. Nevertheless, developing more appropriate QA procedures remains a major task because some parts of them are not applicable to the Na-experiment

  11. An Application Example Analysis of Quality Assurance Program for STELLA(Sodium Integral Effect Test Loop for Safety Simulation and Assessment) Project

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Minhwan; Gam, Dayoung; Eoh, Jae-Hyuk; Jeong, Ji-Young [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    KAERI has been conducting various basic R and D activities in the field of nuclear technology. In addition, KAERI is now participating in the Generation IV International Forum (GIF), preparing for the development of key technologies for Generation IV nuclear energy system, including Sodium cooled Fast Reactor (SFR) development. All of the key technologies for SFR development need an appropriate level of QA activities to achieve the GIF safety and performance objectives. Therefore, QA activities have been conducted as an essential part of the national SFR project. As a result, QAM (Quality Assurance Manual) and QAP (Quality Assurance Procedures) have been developed for the SFR project, which are based on ASME NQA-1, KEPIC QAP and the GIF Quality Management System Guidelines. In this work, the introduction background and application examples of the QA program for the STELLA project were investigated. Application of the QA for the STELLA project has great significance because the QA has been mainly applied for the nuclear power plant area in operation, which helps ensure the reliability of the test data and completeness of the research performance. Nevertheless, developing more appropriate QA procedures remains a major task because some parts of them are not applicable to the Na-experiment.

  12. Mathematical modelling of performance of safety rod and its drive mechanism in sodium cooled fast reactor during scram action

    International Nuclear Information System (INIS)

    Rajan Babu, V.; Thanigaiyarasu, G.; Chellapandi, P.

    2014-01-01

    Highlights: • Mathematical modelling of dynamic behaviour of safety rod during scram action in fast reactor. • Effects of hydraulics, structural interaction and geometry on drop time of safety rod are understood. • Using simplified model, drop time can be assessed replacing detailed CFD analysis. • Sensitivities of the related parameters on drop time are understood. • Experimental validation qualifies the modelling and computer software developed. - Abstract: Performance of safety rod and its drive mechanism which are parts of shutdown systems in sodium cooled fast reactor (SFR) plays a major role in ensuring safe operation of the plant during all the design basis events. The safety rods are to be inserted into the core within a stipulated time during off-normal conditions of the reactor. Mathematical modelling of dynamic behaviour of a safety rod and its drive mechanism in a typical 500 MWe SFR during scram action is considered in the present study. A full-scale prototype system has undergone qualification tests in air, water and in sodium simulating the operating conditions in the reactor. In this paper, the salient features of the safety rod and its mechanism, details related to mathematical modelling and sensitivity of the parameters having influence on drop time are presented. The outcomes of the numerical analysis are compared with the experimental results. In this process, the mathematical model and the computer software developed are validated

  13. Analysis of advanced sodium-cooled fast reactor core designs with improved safety characteristics

    Energy Technology Data Exchange (ETDEWEB)

    Sun, K.

    2012-09-15

    Currently, the large majority of nuclear power plants are operated with thermal-neutron spectra and need regular fuel loading of enriched uranium. According to the identified conventional uranium resources and their current consumption rate, only about 100 years’ nuclear fuel supply is foreseen. A reactor operated with a fast-neutron spectrum, on the other hand, can induce self-sustaining, or even breeding, conditions for its inventory of fissile material, which effectively allow it, after the initial loading, to be refueled using simply natural or depleted uranium. This implies a much more efficient use of uranium resources. Moreover, minor actinides become fissionable in a fast-neutron spectrum, enabling full closure of the fuel cycle and leading to a minimization of long-lived radioactive wastes. The sodium-cooled fast reactor (SFR) is one of the most promising candidates to meet the Generation IV International Forum (GIF) declared goals. In comparison to other Generation IV systems, there is considerable design experience related to the SFR, and also more than 300 reactor years of practical operation. As a fast-neutron-spectrum system, the long-term operation of an SFR core in a closed fuel cycle will lead to an equilibrium state, where both reactivity and fuel mass flow stabilize. Although the SFR has many advantageous characteristics, it has one dominating neutronics drawback: there is generally a positive reactivity effect when sodium coolant is removed from the core. This so-called sodium void effect becomes even stronger in the equilibrium closed fuel cycle. The goal of the present doctoral research is to improve the safety characteristics of advanced SFR core designs, in particular, from the viewpoint of the positive sodium void reactivity effect. In this context, particular importance has been given to the dynamic core behavior under a hypothetical unprotected loss-of-flow (ULOF) accident scenario, in which sodium boiling occurs. The proposed

  14. Analysis of advanced sodium-cooled fast reactor core designs with improved safety characteristics

    International Nuclear Information System (INIS)

    Sun, K.

    2012-09-01

    Currently, the large majority of nuclear power plants are operated with thermal-neutron spectra and need regular fuel loading of enriched uranium. According to the identified conventional uranium resources and their current consumption rate, only about 100 years’ nuclear fuel supply is foreseen. A reactor operated with a fast-neutron spectrum, on the other hand, can induce self-sustaining, or even breeding, conditions for its inventory of fissile material, which effectively allow it, after the initial loading, to be refueled using simply natural or depleted uranium. This implies a much more efficient use of uranium resources. Moreover, minor actinides become fissionable in a fast-neutron spectrum, enabling full closure of the fuel cycle and leading to a minimization of long-lived radioactive wastes. The sodium-cooled fast reactor (SFR) is one of the most promising candidates to meet the Generation IV International Forum (GIF) declared goals. In comparison to other Generation IV systems, there is considerable design experience related to the SFR, and also more than 300 reactor years of practical operation. As a fast-neutron-spectrum system, the long-term operation of an SFR core in a closed fuel cycle will lead to an equilibrium state, where both reactivity and fuel mass flow stabilize. Although the SFR has many advantageous characteristics, it has one dominating neutronics drawback: there is generally a positive reactivity effect when sodium coolant is removed from the core. This so-called sodium void effect becomes even stronger in the equilibrium closed fuel cycle. The goal of the present doctoral research is to improve the safety characteristics of advanced SFR core designs, in particular, from the viewpoint of the positive sodium void reactivity effect. In this context, particular importance has been given to the dynamic core behavior under a hypothetical unprotected loss-of-flow (ULOF) accident scenario, in which sodium boiling occurs. The proposed

  15. The undersea location of the Swedish Final Repository for reactor waste, SFR - human intrusion aspects

    International Nuclear Information System (INIS)

    Eng, T.

    1989-01-01

    The Swedish Final Repository for reactor waste, SFR, is built under the Baltic sea close to the Forsmark nuclear power plant. Sixty metres of rock cover the repository caverns under the seabed. The depth of the Baltic sea is about 5-6 m at this location. A human intrusion scenario that in normal inland locations has shown to be of great importance, is a well that is drilled through or in the close vicinity of the repository. Since the land uplift in the SFR area is about 6 mm/year the undersea location of SFR ensures that no well will be drilled at this location for a considerable time while the area is covered by the Baltic sea

  16. GALAXY STRUCTURE AND MODE OF STAR FORMATION IN THE SFR-MASS PLANE FROM z {approx} 2.5 TO z {approx} 0.1

    Energy Technology Data Exchange (ETDEWEB)

    Wuyts, Stijn; Foerster Schreiber, Natascha M.; Magnelli, Benjamin; Genzel, Reinhard; Lutz, Dieter; Berta, Stefano; Gracia-Carpio, Javier; Nordon, Raanan [Max-Planck-Institut fuer extraterrestrische Physik, Giessenbachstrasse, D-85748 Garching (Germany); Van der Wel, Arjen [Max-Planck-Institut fuer Astronomie, Koenigstuhl 17, D-69117 Heidelberg (Germany); Guo, Yicheng [Astronomy Department, University of Massachusetts, 710 N. Pleasant Street, Amherst, MA 01003 (United States); Aussel, Herve; Le Floc' h, Emeric [Laboratoire AIM, CEA/DSM-CNRS-Universite Paris Diderot, IRFU/Service d' Astrophysique, Bat. 709, CEA-Saclay, F-91191 Gif-sur-Yvette Cedex (France); Barro, Guillermo; Kocevski, Dale D.; McGrath, Elizabeth J. [UCO/Lick Observatory, Department of Astronomy and Astrophysics, University of California, Santa Cruz, CA 95064 (United States); Cava, Antonio [Departamento de Astrofisica, Facultad de CC. Fisicas, Universidad Complutense de Madrid, E-28040 Madrid (Spain); Hathi, Nimish P. [Observatories of the Carnegie Institution of Washington, Pasadena, CA 91101 (United States); Huang, Kuang-Han [Johns Hopkins University, 3400 North Charles Street, Baltimore, MD 21218 (United States); Koekemoer, Anton M. [Space Telescope Science Institute, 3700 San Martin Drive, Baltimore, MD 21218 (United States); Lee, Kyoung-Soo [Yale Center for Astronomy and Astrophysics, Department of Physics, Yale University, New Haven, CT 06520 (United States); and others

    2011-12-01

    We analyze the dependence of galaxy structure (size and Sersic index) and mode of star formation ({Sigma}{sub SFR} and SFR{sub IR}/SFR{sub UV}) on the position of galaxies in the star formation rate (SFR) versus mass diagram. Our sample comprises roughly 640,000 galaxies at z {approx} 0.1, 130,000 galaxies at z {approx} 1, and 36,000 galaxies at z {approx} 2. Structural measurements for all but the z {approx} 0.1 galaxies are based on Hubble Space Telescope imaging, and SFRs are derived using a Herschel-calibrated ladder of SFR indicators. We find that a correlation between the structure and stellar population of galaxies (i.e., a 'Hubble sequence') is already in place since at least z {approx} 2.5. At all epochs, typical star-forming galaxies on the main sequence are well approximated by exponential disks, while the profiles of quiescent galaxies are better described by de Vaucouleurs profiles. In the upper envelope of the main sequence, the relation between the SFR and Sersic index reverses, suggesting a rapid buildup of the central mass concentration in these starbursting outliers. We observe quiescent, moderately and highly star-forming systems to co-exist over an order of magnitude or more in stellar mass. At each mass and redshift, galaxies on the main sequence have the largest size. The rate of size growth correlates with specific SFR, and so does {Sigma}{sub SFR} at each redshift. A simple model using an empirically determined star formation law and metallicity scaling, in combination with an assumed geometry for dust and stars, is able to relate the observed {Sigma}{sub SFR} and SFR{sub IR}/SFR{sub UV}, provided a more patchy dust geometry is assumed for high-redshift galaxies.

  17. GALAXY STRUCTURE AND MODE OF STAR FORMATION IN THE SFR-MASS PLANE FROM z ∼ 2.5 TO z ∼ 0.1

    International Nuclear Information System (INIS)

    Wuyts, Stijn; Förster Schreiber, Natascha M.; Magnelli, Benjamin; Genzel, Reinhard; Lutz, Dieter; Berta, Stefano; Graciá-Carpio, Javier; Nordon, Raanan; Van der Wel, Arjen; Guo, Yicheng; Aussel, Hervé; Le Floc'h, Emeric; Barro, Guillermo; Kocevski, Dale D.; McGrath, Elizabeth J.; Cava, Antonio; Hathi, Nimish P.; Huang, Kuang-Han; Koekemoer, Anton M.; Lee, Kyoung-Soo

    2011-01-01

    We analyze the dependence of galaxy structure (size and Sérsic index) and mode of star formation (Σ SFR and SFR IR /SFR UV ) on the position of galaxies in the star formation rate (SFR) versus mass diagram. Our sample comprises roughly 640,000 galaxies at z ∼ 0.1, 130,000 galaxies at z ∼ 1, and 36,000 galaxies at z ∼ 2. Structural measurements for all but the z ∼ 0.1 galaxies are based on Hubble Space Telescope imaging, and SFRs are derived using a Herschel-calibrated ladder of SFR indicators. We find that a correlation between the structure and stellar population of galaxies (i.e., a 'Hubble sequence') is already in place since at least z ∼ 2.5. At all epochs, typical star-forming galaxies on the main sequence are well approximated by exponential disks, while the profiles of quiescent galaxies are better described by de Vaucouleurs profiles. In the upper envelope of the main sequence, the relation between the SFR and Sérsic index reverses, suggesting a rapid buildup of the central mass concentration in these starbursting outliers. We observe quiescent, moderately and highly star-forming systems to co-exist over an order of magnitude or more in stellar mass. At each mass and redshift, galaxies on the main sequence have the largest size. The rate of size growth correlates with specific SFR, and so does Σ SFR at each redshift. A simple model using an empirically determined star formation law and metallicity scaling, in combination with an assumed geometry for dust and stars, is able to relate the observed Σ SFR and SFR IR /SFR UV , provided a more patchy dust geometry is assumed for high-redshift galaxies.

  18. SDSS IV MaNGA - sSFR profiles and the slow quenching of discs in green valley galaxies

    Science.gov (United States)

    Belfiore, Francesco; Maiolino, Roberto; Bundy, Kevin; Masters, Karen; Bershady, Matthew; Oyarzún, Grecco; Lin, Lihwai; Cano-Diaz, Mariana; Wake, David; Spindler, Ashley; Thomas, Daniel; Brownstein, Joel R.; Drory, Niv; Yan, Renbin

    2018-03-01

    We study radial profiles in Hα equivalent width and specific star formation rate (sSFR) derived from spatially-resolved SDSS-IV MaNGA spectroscopy to gain insight on the physical mechanisms that suppress star formation and determine a galaxy's location in the SFR-M_\\star diagram. Even within the star-forming `main sequence', the measured sSFR decreases with stellar mass, both in an integrated and spatially-resolved sense. Flat sSFR radial profiles are observed for log(M_\\star / M_⊙ ) history. Our primary focus is the green valley, constituted by galaxies lying below the star formation main sequence, but not fully passive. In the green valley we find sSFR profiles that are suppressed with respect to star-forming galaxies of the same mass at all galactocentric distances out to 2 effective radii. The responsible quenching mechanism therefore appears to affect the entire galaxy, not simply an expanding central region. The majority of green valley galaxies of log(M_\\star / M_⊙ ) > 10.0 are classified spectroscopically as central low-ionisation emission-line regions (cLIERs). Despite displaying a higher central stellar mass concentration, the sSFR suppression observed in cLIER galaxies is not simply due to the larger mass of the bulge. Drawing a comparison sample of star forming galaxies with the same M_\\star and Σ _{1 kpc} (the mass surface density within 1 kpc), we show that a high Σ _{1 kpc} is not a sufficient condition for determining central quiescence.

  19. Development of Preliminary HT9 Cladding Tube for Sodium-cooled Fast Reactor (SFR)

    International Nuclear Information System (INIS)

    Kim, Jun Hwan; Baek, Jong Hyuk; Heo, Hyeong Min; Park, Sang Gyu; Kim, Sung Ho; Lee, Chan Bock

    2013-01-01

    To achieve manufacturing technology of the fuel cladding tube in order to keep pace with the predetermined schedule in developing SFR fuel, KAERI has launched in developing fuel cladding tube in cooperation with a domestic steelmaking company. After fabricating medium-sized 1.1 ton HT9 ingot, followed by the multiple processes of hot and cold working, preliminary samples of HT9 seamless cladding tube having 7.4mm in outer diameter, 0.56mm in thickness, and 3m in length were fabricated. The objective of this study is to summarize the brief development status of the HT9 cladding tubes. Mechanical properties like axial tension, biaxial burst, pressurized creep and sodium compatibility of the cladding tubes were carried out to set up the performance evaluation technology to test the prototype FMS cladding tube which is going to be manufactured in next stage. As a part of developing fuel cladding for the Sodium-cooled Fast Reactor (SFR), preliminary HT9 cladding tube was fabricated in cooperation with a domestic steelmaking company. Microstructure as well as mechanical tests like axial tensile test, biaxial burst test, and pressurized creep test of the fuel cladding were carried out. Performance of the domestic HT9 tube was revealed to be similar in the previously fabricated foreign HT9 tube. Further prototype FMS cladding tube is going to be manufactured in next year based on this experience. Various test items like mechanical test, sodium compatibility test, microstructural analysis, basic property, cladding performance under transient situation, and performance under ion and neutron irradiation are going be performed in the future to set up the relevant technology for the licensing of the SFR cladding tube

  20. Development of basic key technologies for Gen IV SFR

    International Nuclear Information System (INIS)

    Kim, Yeongil; Kim, Sungoh; Choi, Sukgi

    2012-04-01

    The advanced concepts, for the breakeven reactor(1,200MWe) and TRU burner(600MWe), were defined to satisfy the technology goals of Generation IV nuclear systems. Based on the advanced design concepts, a conceptual design of the demonstration SFR has been developed using the available licensing technology. The conceptual core design has been developed for the TRU burner in which an initial core is fueled with less than 20wt% enriched U235, and finally transformed to a self-recycled TRU core. The passive decay heat removal circuit ensuring reactor safety even in case of loss of emergency power has been developed and minimization of a reactor vessel and simplification of reactor internals have been conducted in the conceptual design. For development of advanced technologies, control logics for various power levels and the optimal design concept of heat exchanger applicable to supercritical CO 2 Brayton cycle as an energy conversion system was developed. A novel under-sodium waveguide sensor and a prototype under-sodium inspection system have been developed for under-sodium viewing of in-vessel structures submerged in an opaque liquid sodium. The fabrication technology of fuel slugs using the advanced fuel slug casting system was developed, and U-Zr alloy fuel rods were fabricated and examined. And a HT 9 cladding tube was manufactured using the developed cladding tube fabrication technology. For development of basic technologies, the cross section adjustment code ATCROSS and the MATRA-LMR code with HCFs have been developed to reduce core design uncertainties. The SIE ASME-NH computer program to evaluate high temperature structural design for 60 years design life, and the safety analysis code MARS-LMR with thermal-hydraulic and reactivity feedback models have been developed and validated. In addition, the sodium impurity measurement and control technology, the sodium water reaction event propagation model to predict the sodium leak propagation in a steam generator, and

  1. Overview of fourth generation reactors. Assessment in terms of safety and radiation protection

    International Nuclear Information System (INIS)

    Couturier, J.; Baudrand, O.; Blanc, D.; Bourgois, T.; Hache, G.; Ivanov, E.; Bonneville, H.; Meignen, R.; Nicaise, G.; Bruna, G.; Clement, B.; Kissane, M.; Monhardt, B.

    2012-01-01

    Based on a systematic analysis of the different concepts of fourth generation nuclear reactors, this report gives an overview of specific aspects regarding safety and radiation protection for six concepts: sodium fast reactors (SFR), gas fast reactors (GFR), lead fast reactors (LFR), molten salt reactors (MSR), very high or high temperature reactors (V/HTR) and supercritical water reactors (SCWR). This assessment is based on different studies and researches performed by the IRSN at an international level. For each reactor concept, the report proposes a presentation of the current status of development and its perspectives, describes the safety aspects which are specific to this concept, identifies and discusses elements for safety analysis, and assesses the concept with respect to the Fukushima accident and IAEA recommendations and predefined themes

  2. Coupled MCNP - SAS-SFR calculations for sodium fast reactor core at steady-state - 15460

    International Nuclear Information System (INIS)

    Ponomarev, A.; Travleev, A.; Pfrang, W.; Sanchez, V.

    2015-01-01

    The prediction of core parameters at steady state is the first step when studying core accident transient behaviour. At this step thermal hydraulics (TH) and core geometry parameters are calculated corresponding to initial operating conditions. In this study we present the coupling of the SAS-SFR code to the Monte-Carlo neutron transport code MCNP at steady state together with application to the European Sodium Fast Reactor (ESFR). The SAS-SFR code employs a multi-channel core representation where each channel represents subassemblies with similar power, thermal-hydraulics and pin mechanics conditions. For every axial node of every channel the individual geometry and material compositions parameters are calculated in accord with power and cooling conditions. This requires supplying the SAS-SFR-code with nodal power values which should be calculated by neutron physics code with given realistic core parameters. In the conventional approach the neutron physics model employs some core averaged TH and geometry data (fuel temperature, coolant density, core axial and radial expansion). In this study we organize a new approach coupling the MCNP neutron physics models and the SAS-SFR models, so that calculations of power can be improved by using distributed core parameters (TH and geometry) taken from SAS-SFR. The MCNP code is capable to describe cores with distributed TH parameters and even to model non-uniform axial expansion of fuel subassemblies. In this way, core TH and geometrical data calculated by SAS-SFR are taken into account accurately in the neutronics model. The coupling implementation is done by data exchange between two codes with help of processing routines managed by driver routine. Currently it is model-specific and realized for the ESFR 'Reference Oxide' core. The Beginning-Of-Life core state is considered with 10 channel representation for fuel subassemblies. For this model several sets of coupled calculations are performed, in which different

  3. Safety requirements expected to the prototype fast breeder reactor 'Monju'

    International Nuclear Information System (INIS)

    2014-11-01

    In July 2013, Nuclear Regulation Authority (NRA) has enforced new regulatory requirements in consideration of severe accidents for the commercial light water reactors (LWR) and also prototype power generation reactors such as the sodium-cooled fast reactors (SFR) of 'Monju' based on TEPCO Fukushima Daiichi nuclear power plant accident (hereinafter referred to as '1F accident') occurred in March 2011. Although the regulatory requirements for SFR will be revised by NRA with consideration for public comments, Japan Atomic Energy Agency (JAEA) set up 'Advisory Committee on Monju Safety Requirements' consisting of fast breeder reactor (FBR) and safety assessment experts in order to establish original safety requirements expected to the prototype FBR 'Monju' considering severe accidents with knowledge from JAEA as well as scientific and technical insights from the experts. This report summarizes the safety requirements expected to Monju discussed by the committee. (author)

  4. Thermodynamic analysis and preliminary design of closed Brayton cycle using nitrogen as working fluid and coupled to small modular Sodium-cooled fast reactor (SM-SFR)

    International Nuclear Information System (INIS)

    Olumayegun, Olumide; Wang, Meihong; Kelsall, Greg

    2017-01-01

    Highlights: • Nitrogen closed Brayton cycle for small modular sodium-cooled fast reactor studied. • Thermodynamic modelling and analysis of closed Brayton cycle performed. • Two-shaft configuration proposed and performance compared to single shaft. • Preliminary design of heat exchangers and turbomachinery carried out. - Abstract: Sodium-cooled fast reactor (SFR) is considered the most promising of the Generation IV reactors for their near-term demonstration of power generation. Small modular SFRs (SM-SFRs) have less investment risk, can be deployed more quickly, are easier to operate and are more flexible in comparison to large nuclear reactor. Currently, SFRs use the proven Rankine steam cycle as the power conversion system. However, a key challenge is to prevent dangerous sodium-water reaction that could happen in SFR coupled to steam cycle. Nitrogen gas is inert and does not react with sodium. Hence, intercooled closed Brayton cycle (CBC) using nitrogen as working fluid and with a single shaft configuration has been one common power conversion system option for possible near-term demonstration of SFR. In this work, a new two shaft nitrogen CBC with parallel turbines was proposed to further simplify the design of the turbomachinery and reduce turbomachinery size without compromising the cycle efficiency. Furthermore, thermodynamic performance analysis and preliminary design of components were carried out in comparison with a reference single shaft nitrogen cycle. Mathematical models in Matlab were developed for steady state thermodynamic analysis of the cycles and for preliminary design of the heat exchangers, turbines and compressors. Studies were performed to investigate the impact of the recuperator minimum terminal temperature difference (TTD) on the overall cycle efficiency and recuperator size. The effect of turbomachinery efficiencies on the overall cycle efficiency was examined. The results showed that the cycle efficiency of the proposed

  5. Safety balance: Analysis of safety systems

    International Nuclear Information System (INIS)

    Delage, M.; Giroux, C.

    1990-12-01

    Safety analysis, and particularly analysis of exploitation of NPPs is constantly affected by EDF and by the safety authorities and their methodologies. Periodic safety reports ensure that important issues are not missed on daily basis, that incidents are identified and that relevant actions are undertaken. French safety analysis method consists of three principal steps. First type of safety balance is analyzed at the normal start-up phase for each unit including the final safety report. This enables analysis of behaviour of units ten years after their licensing. Second type is periodic operational safety analysis performed during a few years. Finally, the third step consists of safety analysis of the oldest units with the aim to improve the safety standards. The three steps of safety analysis are described in this presentation in detail with the aim to present the objectives and principles. Examples of most recent exercises are included in order to illustrate the importance of such analyses

  6. The Mediator subunit SFR6/MED16 controls defence gene expression mediated by salicylic acid and jasmonate responsive pathways.

    Science.gov (United States)

    Wathugala, Deepthi L; Hemsley, Piers A; Moffat, Caroline S; Cremelie, Pieter; Knight, Marc R; Knight, Heather

    2012-07-01

    • Arabidopsis SENSITIVE TO FREEZING6 (SFR6) controls cold- and drought-inducible gene expression and freezing- and osmotic-stress tolerance. Its identification as a component of the MEDIATOR transcriptional co-activator complex led us to address its involvement in other transcriptional responses. • Gene expression responses to Pseudomonas syringae, ultraviolet-C (UV-C) irradiation, salicylic acid (SA) and jasmonic acid (JA) were investigated in three sfr6 mutant alleles by quantitative real-time PCR and susceptibility to UV-C irradiation and Pseudomonas infection were assessed. • sfr6 mutants were more susceptible to both Pseudomonas syringae infection and UV-C irradiation. They exhibited correspondingly weaker PR (pathogenesis-related) gene expression than wild-type Arabidopsis following these treatments or after direct application of SA, involved in response to both UV-C and Pseudomonas infection. Other genes, however, were induced normally in the mutants by these treatments. sfr6 mutants were severely defective in expression of plant defensin genes in response to JA; ectopic expression of defensin genes was provoked in wild-type but not sfr6 by overexpression of ERF5. • SFR6/MED16 controls both SA- and JA-mediated defence gene expression and is necessary for tolerance of Pseudomonas syringae infection and UV-C irradiation. It is not, however, a universal regulator of stress gene transcription and is likely to mediate transcriptional activation of specific regulons only. © 2012 The Authors. New Phytologist © 2012 New Phytologist Trust.

  7. Swi5-Sfr1 protein stimulates Rad51-mediated DNA strand exchange reaction through organization of DNA bases in the presynaptic filament.

    KAUST Repository

    Fornander, Louise H

    2013-12-03

    The Swi5-Sfr1 heterodimer protein stimulates the Rad51-promoted DNA strand exchange reaction, a crucial step in homologous recombination. To clarify how this accessory protein acts on the strand exchange reaction, we have analyzed how the structure of the primary reaction intermediate, the Rad51/single-stranded DNA (ssDNA) complex filament formed in the presence of ATP, is affected by Swi5-Sfr1. Using flow linear dichroism spectroscopy, we observe that the nucleobases of the ssDNA are more perpendicularly aligned to the filament axis in the presence of Swi5-Sfr1, whereas the bases are more randomly oriented in the absence of Swi5-Sfr1. When using a modified version of the natural protein where the N-terminal part of Sfr1 is deleted, which has no affinity for DNA but maintained ability to stimulate the strand exchange reaction, we still observe the improved perpendicular DNA base orientation. This indicates that Swi5-Sfr1 exerts its activating effect through interaction with the Rad51 filament mainly and not with the DNA. We propose that the role of a coplanar alignment of nucleobases induced by Swi5-Sfr1 in the presynaptic Rad51/ssDNA complex is to facilitate the critical matching with an invading double-stranded DNA, hence stimulating the strand exchange reaction.

  8. Quality Assurance Program Plan for SFR Metallic Fuel Data Qualification

    Energy Technology Data Exchange (ETDEWEB)

    Benoit, Timothy [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Hlotke, John Daniel [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Yacout, Abdellatif [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division

    2017-07-05

    This document contains an evaluation of the applicability of the current Quality Assurance Standards from the American Society of Mechanical Engineers Standard NQA-1 (NQA-1) criteria and identifies and describes the quality assurance process(es) by which attributes of historical, analytical, and other data associated with sodium-cooled fast reactor [SFR] metallic fuel and/or related reactor fuel designs and constituency will be evaluated. This process is being instituted to facilitate validation of data to the extent that such data may be used to support future licensing efforts associated with advanced reactor designs. The initial data to be evaluated under this program were generated during the US Integral Fast Reactor program between 1984-1994, where the data includes, but is not limited to, research and development data and associated documents, test plans and associated protocols, operations and test data, technical reports, and information associated with past United States Nuclear Regulatory Commission reviews of SFR designs.

  9. Safety margins in deterministic safety analysis

    International Nuclear Information System (INIS)

    Viktorov, A.

    2011-01-01

    The concept of safety margins has acquired certain prominence in the attempts to demonstrate quantitatively the level of the nuclear power plant safety by means of deterministic analysis, especially when considering impacts from plant ageing and discovery issues. A number of international or industry publications exist that discuss various applications and interpretations of safety margins. The objective of this presentation is to bring together and examine in some detail, from the regulatory point of view, the safety margins that relate to deterministic safety analysis. In this paper, definitions of various safety margins are presented and discussed along with the regulatory expectations for them. Interrelationships of analysis input and output parameters with corresponding limits are explored. It is shown that the overall safety margin is composed of several components each having different origins and potential uses; in particular, margins associated with analysis output parameters are contrasted with margins linked to the analysis input. While these are separate, it is possible to influence output margins through the analysis input, and analysis method. Preserving safety margins is tantamount to maintaining safety. At the same time, efficiency of operation requires optimization of safety margins taking into account various technical and regulatory considerations. For this, basic definitions and rules for safety margins must be first established. (author)

  10. Safety analysis fundamentals

    International Nuclear Information System (INIS)

    Wright, A.C.D.

    2002-01-01

    This paper discusses the safety analysis fundamentals in reactor design. This study includes safety analysis done to show consequences of postulated accidents are acceptable. Safety analysis is also used to set design of special safety systems and includes design assist analysis to support conceptual design. safety analysis is necessary for licensing a reactor, to maintain an operating license, support changes in plant operations

  11. Comparison of Design Concepts for SFR under Development

    Energy Technology Data Exchange (ETDEWEB)

    Suh, Namduk; Choi, Yongwon; Bae, Moohoon; Shin, Andong [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2016-10-15

    The goal of ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) with a capacity of 600 MWe is to study the technical demonstration that can be scaled up to commercial reactor. It was expected that the success of ASTRID project could eventually lead to operation of industrial reactor around 2040. On 2012, ASTRID designer has submitted the DOrS (Dossier d’Orientations de Sûreté, Safety Orientation Document) for ASTRID to IRSN and IRSN has issued a report after reviewing the DOrS. The report DOrS itself is not available publicly, intellectual property might be the reason, but the review document of IRSN is open to public, so we can understand the basic concept of ASTRID by IRSN report. The DOrS of ASTRID and the TTR for PGSFR have not the same format and also the same purpose, so it is not easy to compare the two design concepts directly. But, still, we think the concepts could be compared in a very general way. Thus, in this paper we have presented the very short comparison results of the two SFR design. Our opinion after first reviewing the TTR is that the PGSFR needs to be designed in a more systematic way. The requirements are coming basically from the previous document used for SMART licensing and do not show prototype reactor specific characters.

  12. An Overview and Evaluation of U.S. SFR Safety Approach

    International Nuclear Information System (INIS)

    Sofu, Tanju

    2013-01-01

    Safety Approach: • The traditional approach to demonstrating adequacy of defense-in-depth in a design is deterministic, but a combination of deterministic and probabilistic approaches is increasingly recommended for especially for advanced reactors. – Deterministic approach classifies initiating events by likelihood, while the risk-informed approach introduces a quantified probability estimate. • Risk-informed and performance-based safety approach considers both probability and consequences of events. – Accidents with large consequences are reduced in risk significance by requiring that their probability are acceptably small

  13. Experimental results of passive vibro-acoustic leak detection in SFR steam generator mock-up

    International Nuclear Information System (INIS)

    Moriot, J.; Gastaldi, O.; Maxit, L.; Guyader, J-L.; Perisse, J.; Migot, B.

    2013-06-01

    Regarding to GEN 4 context, it is necessary to fulfil the high safety standards for sodium fast reactors (SFR), particularly against water-sodium reaction which may occur in the steam generator units (SGU) in case of leak. This reaction can cause severe damages in the component in a short time. Detecting such a leak by visual in-sodium inspection is impossible because of sodium opacity. Hydrogen detection is then used but the time response of this method can be high in certain operating conditions. Active and passive acoustic leak detection methods were studied before SUPERPHENIX plant shutdown in 1997 to detect a water-into-sodium leak with a short time response. In the context of the new R and D studies for SFR, an innovative passive vibro-acoustic method is developed in the framework of a Ph.D. thesis to match with GEN 4 safety requirements. The method consists in assuming that a small leak emits spherical acoustic waves in a broadband frequency domain, which propagate in the liquid sodium and excite the SGU cylindrical shell. These spatially coherent waves are supposed to be buried by a spatially incoherent background noise. The radial velocities of the shell is measured by an array of accelerometers positioned on the external envelop of the SGU and a beam forming treatment is applied to increase the signal-to-noise ratio (SNR) and to detect and localize the acoustic source. Previous numerical experiments were achieved and promising results were obtained. In this paper, experimental results of the proposed passive vibro-acoustic leak detection are presented. The experiment consists in a cylindrical water-filled steel pipe representing a model of SGU shell without tube bundle. A hydro-phone emitting an acoustic signal is used to simulate an acoustic monopole. Spatially uncorrelated noise or water-flow induced shell vibrations are considered as the background noise. The beam-forming method is applied to vibration signals measured by a linear array of

  14. A New Streamflow-Routing (SFR1) Package to Simulate Stream-Aquifer Interaction with MODFLOW-2000

    Science.gov (United States)

    Prudic, David E.; Konikow, Leonard F.; Banta, Edward R.

    2004-01-01

    The increasing concern for water and its quality require improved methods to evaluate the interaction between streams and aquifers and the strong influence that streams can have on the flow and transport of contaminants through many aquifers. For this reason, a new Streamflow-Routing (SFR1) Package was written for use with the U.S. Geological Survey's MODFLOW-2000 ground-water flow model. The SFR1 Package is linked to the Lake (LAK3) Package, and both have been integrated with the Ground-Water Transport (GWT) Process of MODFLOW-2000 (MODFLOW-GWT). SFR1 replaces the previous Stream (STR1) Package, with the most important difference being that stream depth is computed at the midpoint of each reach instead of at the beginning of each reach, as was done in the original Stream Package. This approach allows for the addition and subtraction of water from runoff, precipitation, and evapotranspiration within each reach. Because the SFR1 Package computes stream depth differently than that for the original package, a different name was used to distinguish it from the original Stream (STR1) Package. The SFR1 Package has five options for simulating stream depth and four options for computing diversions from a stream. The options for computing stream depth are: a specified value; Manning's equation (using a wide rectangular channel or an eight-point cross section); a power equation; or a table of values that relate flow to depth and width. Each stream segment can have a different option. Outflow from lakes can be computed using the same options. Because the wetted perimeter is computed for the eight-point cross section and width is computed for the power equation and table of values, the streambed conductance term no longer needs to be calculated externally whenever the area of streambed changes as a function of flow. The concentration of solute is computed in a stream network when MODFLOW-GWT is used in conjunction with the SFR1 Package. The concentration of a solute in a

  15. Overview of nuclear safety activities performed by JRC-IE on Gen IV fast reactor concepts

    Energy Technology Data Exchange (ETDEWEB)

    Tsige-Tamirat, H.; Ammirabile, L.; D' Agata, E.; Fuetterer, M.; Ranguelova, V. [European Commission, Joint Research Centre, Institute for Energy, Westerduinweg 3, 1755LE Petten (Netherlands)

    2010-07-01

    The European Strategic Energy Technology (SET) Plan recognizes the need to develop new energy technologies, in order to reduce greenhouse gas emissions and secure energy supply in Europe. Besides renewable energy and improved energy efficiency, a new generation of nuclear power plants and innovative nuclear power applications can play a significant role to achieve this goal. The JRC Institute for Energy 'Safety of Future Nuclear Reactors' (SFNR) Unit is engaged in experimental research, numerical simulation and modelling, scientific, feasibility and engineering studies on innovative nuclear reactor systems. This also represents a significant EURATOM contribution to the Generation IV International Forum. Its activities deal with, among others, the performance assessment of innovative fuels and materials, development of new reactor core concepts and safety solutions, and knowledge management and preservation. Special attention is given to fast reactor concepts, namely the sodium (SFR) and lead (LFR) cooled reactors. Recognizing the maturity of the SFR technology, the European Sustainable Nuclear Energy Technology Platform (SNETP) considers a prototype SFR to be built as a next-step towards the deployment of a first-of-a-kind Gen IV SFR. This paper gives an overview of current research preformed at JRC-IE with emphasis on the work performed in the Collaborative Project on European Sodium Fast Reactor (CP-ESFR) within the European Commission's Seventh Framework Program. (authors)

  16. Overview of nuclear safety activities performed by JRC-IE on Gen IV fast reactor concepts

    International Nuclear Information System (INIS)

    Tsige-Tamirat, H.; Ammirabile, L.; D'Agata, E.; Fuetterer, M.; Ranguelova, V.

    2010-01-01

    The European Strategic Energy Technology (SET) Plan recognizes the need to develop new energy technologies, in order to reduce greenhouse gas emissions and secure energy supply in Europe. Besides renewable energy and improved energy efficiency, a new generation of nuclear power plants and innovative nuclear power applications can play a significant role to achieve this goal. The JRC Institute for Energy 'Safety of Future Nuclear Reactors' (SFNR) Unit is engaged in experimental research, numerical simulation and modelling, scientific, feasibility and engineering studies on innovative nuclear reactor systems. This also represents a significant EURATOM contribution to the Generation IV International Forum. Its activities deal with, among others, the performance assessment of innovative fuels and materials, development of new reactor core concepts and safety solutions, and knowledge management and preservation. Special attention is given to fast reactor concepts, namely the sodium (SFR) and lead (LFR) cooled reactors. Recognizing the maturity of the SFR technology, the European Sustainable Nuclear Energy Technology Platform (SNETP) considers a prototype SFR to be built as a next-step towards the deployment of a first-of-a-kind Gen IV SFR. This paper gives an overview of current research preformed at JRC-IE with emphasis on the work performed in the Collaborative Project on European Sodium Fast Reactor (CP-ESFR) within the European Commission's Seventh Framework Program. (authors)

  17. Third Joint GIF–IAEA Workshop on Safety Design Criteria for Sodium-Cooled Fast Reactors, 26-27 February 2013, Vienna, Austria. Summary Report

    International Nuclear Information System (INIS)

    2013-01-01

    The main objectives of the meeting were to: • Present and share information on the work carried out by GIF, the IAEA and the Member States on the definition of safety design criteria for SFR, including safety approach and requirements on general plant design; • Present the document prepared by the GIF-SFR Task Force on Safety Design Criteria; • Present and discuss safety design concepts of SFRs under development in Member States, with particular emphasis on design measures against Design Basis Accidents and Design Extended Conditions, as well as the associated safety evaluations and supporting R&D; • Draft a room document which should be the basis of the discussion for the Panel on Safety Design Criteria of the FR13 Conference in Paris. • Discuss the results and agree on the future actions of the 3rd Joint GIF-IAEA Workshop on Safety of Sodium-Cooled Fast Reactors

  18. Development of electromagnetic acoustic transducer (EMAT) phased arrays for SFR inspection

    Energy Technology Data Exchange (ETDEWEB)

    Le Bourdais, Florian; Marchand, Benoît [CEA LIST, Centre de Saclay F-91191 Gif-sur-Yvette (France)

    2014-02-18

    A long-standing problem for Sodium cooled Fast Reactor (SFR) instrumentation is the development of efficient under-sodium visualization systems adapted to the hot and opaque sodium environment. Electromagnetic Acoustic Transducers (EMAT) are potential candidates for a new generation of Ultrasonic Testing (UT) probes well-suited for SFR inspection that can overcome drawbacks of classical piezoelectric probes in sodium environment. Based on the use of new CIVA simulation tools, we have designed and optimized an advanced EMAT probe for under-sodium visualization. This has led to the development of a fully functional L-wave EMAT sensing system composed of 8 elements and a casing withstanding 200° C sodium inspection. Laboratory experiments demonstrated the probe's ability to sweep an ultrasonic beam to an angle of 15 degrees. Testing in a specialized sodium facility has shown that it was possible to obtain pulse-echo signals from a target under several different angles from a fixed position.

  19. Performance Estimation of Supercritical CO2 Cycle for the PG-SFR application with Heat Sink Temperature Variation

    International Nuclear Information System (INIS)

    Ahn, Yoonhan; Cho, Seong Kuk; Lee, Jeong Ik

    2015-01-01

    The heat sink temperature conditions are referred from the annual database of sea water temperature in East sea. When the heat sink temperature increases, the compressor inlet temperature can be influenced and the sudden power decrease can happen due to the large water pumping power. When designing the water pump, the pumping margin should be considered as well. As a part of Prototype Generation IV Sodium-cooled Fast Reactor (PG-SFR) development, the Supercritical CO 2 cycle (S-CO 2 ) is considered as one of the promising candidate that can potentially replace the steam Rankine cycle. S-CO 2 cycle can achieve distinctively high efficiency compared to other Brayton cycles and even competitive performance to the steam Rankine cycle under the mild turbine inlet temperature region. Previous studies explored the optimum size of the S-CO 2 cycle considering component designs including turbomachinery, heat exchangers and pipes. Based on the preliminary design, the thermal efficiency is 31.5% when CO 2 is sufficiently cooled to the design temperature. However, the S-CO 2 compressor performance is highly influenced by the inlet temperature and the compressor inlet temperature can be changed when the heat sink temperature, in this case sea water temperature varies. To estimate the S-CO 2 cycle performance of PG-SFR in the various regions, a Quasi-static system analysis code for S-CO 2 cycle is developed by the KAIST research team. A S-CO 2 cycle for PG-SFR is designed and assessed for off-design performance with the heat sink temperature variation

  20. PGSFR Core Thermal Design Procedure to Evaluate the Safety Margin

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Sun Rock; Kim, Sang-Ji [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    The Korea Atomic Energy Research Institute (KAERI) has performed a SFR design with the final goal of constructing a prototype plant by 2028. The main objective of the SFR prototype plant is to verify the TRU metal fuel performance, reactor operation, and transmutation ability of high-level wastes. The core thermal design is to ensure the safe fuel performance during the whole plant operation. Compared to the critical heat flux in typical light water reactors, nuclear fuel damage in SFR subassemblies arises from a creep induced failure. The creep limit is evaluated based on the maximum cladding temperature, power, neutron flux, and uncertainties in the design parameters, as shown in Fig. 1. In this work, the core thermal design procedures are compared to verify the present PGSFR methodology based on the nuclear plant design criteria/guidelines and previous SFR thermal design methods. The PGSFR core thermal design procedure is verified based on the nuclear plant design criteria/guidelines and previous methods in LWRs and SFRs. The present method aims to directly evaluate the fuel cladding failure and to assure more safety margin. The 2 uncertainty is similar to 95% one-side tolerance limit of 1.96 in LWRs. The HCFs, ITDP, and MCM reveal similar uncertainty propagation for cladding midwall temperature for typical SFR conditions. The present HCFs are mainly employed from the CRBR except the fuel-related uncertainty such as an incorrect fuel distribution. Preliminary PGSFR specific HCFs will be developed by the end of 2015.

  1. Sodium Fast Reactor Safety and Licensing Research Plan

    International Nuclear Information System (INIS)

    Denman, M.; Lachance, J.; Sofu, T.; Bari, R.; Flanagon, G.; Wigeland, R.

    2015-01-01

    This paper summarizes potential research priorities for the US Department of Energy (DOE) with the intent of improving the licensability of the sodium cooled fast reactor (SFR). In support of this project, five panels were tasked with identifying potential safety related gaps in the available information, data and models needed to support the licensing of an SFR. The areas examined were sodium technology; accident sequences and initiators; source term characterization, codes and methods; and fuels and materials. It is the intent of this paper to utilize a structured and transparent process that incorporates feedback from all interested stakeholders to suggest future funding priorities for SFR research and development. While numerous gaps were identified, two cross-cutting gaps related to knowledge preservation were agreed upon by all panels and should be addressed in the near future. The first gap is a need to re-evaluate the current procedures for removing the applied technology access control designation from old documents. The second cross-cutting gap is the need for a robust knowledge management and preservation system in all SFR research areas. Closure of these and the other identified gaps will require both a reprioritization of funding within DOE as well as a re-evaluation of existing bureaucratic procedures within the DOE associated with applied technology and knowledge management. (author)

  2. Site investigation SFR. Rock type coding, overview geological mapping and identification of rock units and possible deformation zones in drill cores from the construction of SFR

    Energy Technology Data Exchange (ETDEWEB)

    Petersson, Jesper (Vattenfall Power Consultant AB, Stockholm (Sweden)); Curtis, Philip; Bockgaard, Niclas (Golder Associates AB (Sweden)); Mattsson, Haakan (GeoVista AB, Luleaa (Sweden))

    2011-01-15

    This report presents the rock type coding, overview lithological mapping and identification of rock units and possible deformation zones in drill cores from 32 boreholes associated with the construction of SFR. This work can be seen as complementary to single-hole interpretations of other older SFR boreholes earlier reported in /Petersson and Andersson 2010/: KFR04, KFR08, KFR09, KFR13, KFR35, KFR36, KFR54, KFR55, KFR7A, KFR7B and KFR7C. Due to deficiencies in the available material, the necessary activities have deviated somewhat from the established methodologies used during the recent Forsmark site investigations for the final repository for spent nuclear fuel. The aim of the current work has been, wherever possible, to allow the incorporation of all relevant material from older boreholes in the ongoing SFR geological modelling work in spite of the deficiencies. The activities include: - Rock type coding of the original geological mapping according to the nomenclature used during the preceding Forsmark site investigation. As part of the Forsmark site investigation such rock type coding has already been performed on most of the old SFR boreholes if the original geological mapping results were available. This earlier work has been complemented by rock type coding on two further boreholes: KFR01 and KFR02. - Lithological overview mapping, including documentation of (1) rock types, (2) ductile and brittle-ductile deformation and (3) alteration for drill cores from eleven of the boreholes for which no original geological borehole mapping was available (KFR31, KFR32, KFR34, KFR37,KFR38, KFR51, KFR69, KFR70, KFR71, KFR72 and KFR89). - Identification of possible deformation zones and merging of similar rock types into rock units. This follows SKB's established criteria and methodology of the geological Single-hole interpretation (SHI) process wherever possible. Deviations from the standard SHI process are associated with the lack of data, for example BIPS images

  3. Site investigation SFR. Rock type coding, overview geological mapping and identification of rock units and possible deformation zones in drill cores from the construction of SFR

    International Nuclear Information System (INIS)

    Petersson, Jesper; Curtis, Philip; Bockgaard, Niclas; Mattsson, Haakan

    2011-01-01

    This report presents the rock type coding, overview lithological mapping and identification of rock units and possible deformation zones in drill cores from 32 boreholes associated with the construction of SFR. This work can be seen as complementary to single-hole interpretations of other older SFR boreholes earlier reported in /Petersson and Andersson 2010/: KFR04, KFR08, KFR09, KFR13, KFR35, KFR36, KFR54, KFR55, KFR7A, KFR7B and KFR7C. Due to deficiencies in the available material, the necessary activities have deviated somewhat from the established methodologies used during the recent Forsmark site investigations for the final repository for spent nuclear fuel. The aim of the current work has been, wherever possible, to allow the incorporation of all relevant material from older boreholes in the ongoing SFR geological modelling work in spite of the deficiencies. The activities include: - Rock type coding of the original geological mapping according to the nomenclature used during the preceding Forsmark site investigation. As part of the Forsmark site investigation such rock type coding has already been performed on most of the old SFR boreholes if the original geological mapping results were available. This earlier work has been complemented by rock type coding on two further boreholes: KFR01 and KFR02. - Lithological overview mapping, including documentation of (1) rock types, (2) ductile and brittle-ductile deformation and (3) alteration for drill cores from eleven of the boreholes for which no original geological borehole mapping was available (KFR31, KFR32, KFR34, KFR37,KFR38, KFR51, KFR69, KFR70, KFR71, KFR72 and KFR89). - Identification of possible deformation zones and merging of similar rock types into rock units. This follows SKB's established criteria and methodology of the geological Single-hole interpretation (SHI) process wherever possible. Deviations from the standard SHI process are associated with the lack of data, for example BIPS images, or a

  4. Sodium fast reactor safety and licensing research plan. Volume I.

    Energy Technology Data Exchange (ETDEWEB)

    Sofu, Tanju (Argonne National Laboratory, Argonne, IL); LaChance, Jeffrey L.; Bari, R. (Brokhaven National Laboratory Upton, NY); Wigeland, Roald (Idaho National Laboratory, Idaho Falls, ID); Denman, Matthew R.; Flanagan, George F. (Oak Ridge National Laboratory, Oak Ridge, TN)

    2012-05-01

    This report proposes potential research priorities for the Department of Energy (DOE) with the intent of improving the licensability of the Sodium Fast Reactor (SFR). In support of this project, five panels were tasked with identifying potential safety-related gaps in available information, data, and models needed to support the licensing of a SFR. The areas examined were sodium technology, accident sequences and initiators, source term characterization, codes and methods, and fuels and materials. It is the intent of this report to utilize a structured and transparent process that incorporates feedback from all interested stakeholders to suggest future funding priorities for the SFR research and development. While numerous gaps were identified, two cross-cutting gaps related to knowledge preservation were agreed upon by all panels and should be addressed in the near future. The first gap is a need to re-evaluate the current procedures for removing the Applied Technology designation from old documents. The second cross-cutting gap is the need for a robust Knowledge Management and Preservation system in all SFR research areas. Closure of these and the other identified gaps will require both a reprioritization of funding within DOE as well as a re-evaluation of existing bureaucratic procedures within the DOE associated with Applied Technology and Knowledge Management.

  5. Sodium fast reactor safety and licensing research plan - Volume I

    International Nuclear Information System (INIS)

    Sofu, Tanju; LaChance, Jeffrey L.; Bari, R.; Wigeland, Roald; Denman, Matthew R.; Flanagan, George F.

    2012-01-01

    This report proposes potential research priorities for the Department of Energy (DOE) with the intent of improving the licensability of the Sodium Fast Reactor (SFR). In support of this project, five panels were tasked with identifying potential safety-related gaps in available information, data, and models needed to support the licensing of a SFR. The areas examined were sodium technology, accident sequences and initiators, source term characterization, codes and methods, and fuels and materials. It is the intent of this report to utilize a structured and transparent process that incorporates feedback from all interested stakeholders to suggest future funding priorities for the SFR research and development. While numerous gaps were identified, two cross-cutting gaps related to knowledge preservation were agreed upon by all panels and should be addressed in the near future. The first gap is a need to re-evaluate the current procedures for removing the Applied Technology designation from old documents. The second cross-cutting gap is the need for a robust Knowledge Management and Preservation system in all SFR research areas. Closure of these and the other identified gaps will require both a reprioritization of funding within DOE as well as a re-evaluation of existing bureaucratic procedures within the DOE associated with Applied Technology and Knowledge Management.

  6. Future extension of the Swedish repository for low and intermediate level waste (SFR)

    International Nuclear Information System (INIS)

    Carlsson, Jan

    2006-01-01

    The existing Swedish repository for low and intermediate level waste (SFR) is licensed for disposal of short-lived waste originated from operation and maintenance of Swedish nuclear power plants. The repository is foreseen to be extended to accommodate short-lived waste from the future decommissioning of the Nuclear Power Plants. Long-lived waste from operation, maintenance and eventually decommissioning will be stored some years before disposal in a geological repository. This repository can be build either as a further extension of the SFR facility or as a separate repository. This paper discusses the strategy of a step-wise extended repository where the extensions are performed during operation of the existing parts of the repository. It describes the process for licensing new parts of the repository (and re-license of the existing parts). (author)

  7. Guide on Project Web Access of SFR R and D and Technology Monitoring System

    International Nuclear Information System (INIS)

    Lee, Dong Uk; Won, Byung Chool; Lee, Yong Bum; Kim, Young In; Hahn, Do Hee

    2008-09-01

    The SFR R and D and technology monitoring system based on the MS enterprise project management is developed for systematic effective management of 'Development of Basic Key Technologies for Gen IV SFR' project which was performed under the Mid- and Long-term Nuclear R and D Program sponsored by the Ministry of Education, Science and Technology. This system is a tool for project management based on web access. Therefore this manual is a detailed guide for Project Web Access(PWA). Section 1 describes the common guide for using of system functions such as project server 2007 client connection setting, additional outlook function setting etc. The section 2 describes the guide for system administrator. It is described the guide for project management in section 3, 4

  8. Guide on Project Web Access of SFR R and D and Technology Monitoring System

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dong Uk; Won, Byung Chool; Lee, Yong Bum; Kim, Young In; Hahn, Do Hee

    2008-09-15

    The SFR R and D and technology monitoring system based on the MS enterprise project management is developed for systematic effective management of 'Development of Basic Key Technologies for Gen IV SFR' project which was performed under the Mid- and Long-term Nuclear R and D Program sponsored by the Ministry of Education, Science and Technology. This system is a tool for project management based on web access. Therefore this manual is a detailed guide for Project Web Access(PWA). Section 1 describes the common guide for using of system functions such as project server 2007 client connection setting, additional outlook function setting etc. The section 2 describes the guide for system administrator. It is described the guide for project management in section 3, 4.

  9. Establishment of Experimental Apparatus and Mechanical Test for SFR Metallic Fuel

    International Nuclear Information System (INIS)

    Kim, Sun Ki; Lee, Chong Tak; Oh, Seok Jin; Ko, Young Mo; Kim, Ki Hwan; Woo, Yoon Myung; Lee, Chan Bock

    2010-12-01

    U-Zr binary alloys and U-Zr-Ce ternary alloys as SFR surrogate metallic fuels were fabricated by a casting process. Tensile tests were performed to evaluate the mechanical properties of the fuels. As a results, the mechanical properties such as yield strength, ultimate tensile strength, and elongation were measured. In this report, these experimental results are presented

  10. DBE Analysis for KALIMER-600

    International Nuclear Information System (INIS)

    Ha, Kwi Seok; Jeong, Hae Young; Kwon, Young Min; Chang, Won Pyo; Lee, Yong Bum; Kim, Young II

    2009-01-01

    The SFR (Sodium Fast Reactor) which is being developed at KAERI (Korea Atomic Energy Research Institute) is currently divided into three types, such as, Advanced Concept 600 MWe break-even reactor and burner reactor and 1200 MWe break-even reactor. As a part of accidents analysis of the 600 MWe break-even reactor, 5 representative DBE's (Design Bases Events) are analyzed for the safety analysis. The 5 DBE's are TOP (Transient of Over Power), LOF (Loss Of Flow), LOHS (Loss Of Heat Sink), Pipe Break, and SBO (Station Black Out)

  11. Sodium Fast Reactor Safety and Licensing Research Plan

    International Nuclear Information System (INIS)

    Denman, Matthew; Lachance, Jeff; Sofu, Tanju; Wigeland, Roald; Flanagan, George; Bari, Robert

    2013-01-01

    Conclusions: The Sodium Fast Reactor Safety and Licensing Research Plan reports conclude a multi-year expert elicitation process. All information included in the studies are publicly available and the reports are UUR. These reports are intended to guide SFR researchers in the safety and licensing arena to important and outstanding issues Two (and a half) projects have been funded based on the recommendations in this report: • Modernization of SAS4A; • Incorporation of Contain/LMR with MELCOR; • (Data recovery at INL and PNNL)

  12. Linking Safety Analysis to Safety Requirements

    DEFF Research Database (Denmark)

    Hansen, Kirsten Mark

    Software for safety critical systems must deal with the hazards identified by safety analysistechniques: Fault trees, event trees,and cause consequence diagrams can be interpreted as safety requirements and used in the design activity. We propose that the safety analysis and the system design use...

  13. Trade-off study on the power capacity of a prototype SFR in Korea

    International Nuclear Information System (INIS)

    Baek, M. H.; Kim, S. J.; Yoo, J.; Bae, I. H.

    2012-01-01

    The major roles of a prototype SFR are to provide irradiation test capability for the fuel and structure materials, and to obtain operational experiences of systems. Due to a compromise between the irradiation capability and construction costs, the power level should be properly determined. In this paper, a trade-off study on the power level of the prototype SFR was performed from a neutronics viewpoint. To select candidate cores, the parametric study of pin diameters was estimated using 20 wt.% uranium fuel. The candidate cores of different power levels, 125 MWt, 250 MWt, 400 MWt, and 500 MWt, were compared with the 1500 MWt reference core. The resulting core performance and economic efficiency indices became insensitive to the power at about 400-500 MWt and sharply deteriorated at about 125-250 MWt with decreasing core sizes. Fuel management scheme, TRU core performance comparing with uranium core, and sodium void reactivity were also evaluated with increasing power levels. It is found that increasing the number of batches showed higher burnup performance and economic efficiency. However, increasing the cycle length showed the trends in lower economic efficiency. Irradiation performance of TRU and enriched TRU cores was improved about 20 % and 50 %, respectively. The maximum sodium void reactivity of 5.2$ was confirmed less than the design limit of 7.5$. As a result, the power capacity of the prototype SFR should not be less than 250 MWt and would be appropriate at ∼ 500 MWt considering the performance and economic efficiency. (authors)

  14. Fabrication of U-10wt.%Zr Fuel slug for SFR by Injection Casting

    International Nuclear Information System (INIS)

    Kim, Jong Hwan; Song, Hoon; Kim, Hyung Tae; Ko, Young Mo; Kim, Ki Hwan; Lee, Chan B.

    2013-01-01

    The fabrication technology of metal fuel has been developed by various methods such as rolling, swaging, wire drawing, and co-extrusion, but each of these methods had process limitations requiring an additional subsequent process, and needing the fabrication equipment is complex, which is not favorable for remote use. A practical process of metallic fuel fabrication for an SFR needs to be cost efficient, suitable for remote operation, and capable of mass production while reducing the amount of radioactive waste. Injection casting was chosen as the most promising technique, in the early 1950s, and this technique has been applied to fuel slug fabrication for the Experimental Breeder Reactor-II (EBR-II) driver and the Fast Flux Test Facility (FFTF) fuel pins. Because of the simplistic nature of the process and equipment, compared to other processes examined, this process has been successfully used in a remote operation environment for fueling of the EBR-II reactor. In this study, vacuum injection casting suitable for remote operation has been developed to fabricate metallic fuel for an SFR. Vacuum injection casting technique was developed to fabricate metallic fuel for an SFR. The appearance of the fabricated U-10wt.%Zr fuel was generally sound and the internal integrity was found to be satisfactory through gamma-ray radiography. Minimum fuel losses after casting relative to the initial charge amount of U-10wt.%Zr fuel slugs met the proposed goal of less than 0.1% fuel losses during fabrication. Modifications of the current facility system and advanced casting techniques are underway to produce higher quality fuel slugs

  15. Fabrication of U-10wt.%Zr Fuel slug for SFR by Injection Casting

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jong Hwan; Song, Hoon; Kim, Hyung Tae; Ko, Young Mo; Kim, Ki Hwan; Lee, Chan B. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    The fabrication technology of metal fuel has been developed by various methods such as rolling, swaging, wire drawing, and co-extrusion, but each of these methods had process limitations requiring an additional subsequent process, and needing the fabrication equipment is complex, which is not favorable for remote use. A practical process of metallic fuel fabrication for an SFR needs to be cost efficient, suitable for remote operation, and capable of mass production while reducing the amount of radioactive waste. Injection casting was chosen as the most promising technique, in the early 1950s, and this technique has been applied to fuel slug fabrication for the Experimental Breeder Reactor-II (EBR-II) driver and the Fast Flux Test Facility (FFTF) fuel pins. Because of the simplistic nature of the process and equipment, compared to other processes examined, this process has been successfully used in a remote operation environment for fueling of the EBR-II reactor. In this study, vacuum injection casting suitable for remote operation has been developed to fabricate metallic fuel for an SFR. Vacuum injection casting technique was developed to fabricate metallic fuel for an SFR. The appearance of the fabricated U-10wt.%Zr fuel was generally sound and the internal integrity was found to be satisfactory through gamma-ray radiography. Minimum fuel losses after casting relative to the initial charge amount of U-10wt.%Zr fuel slugs met the proposed goal of less than 0.1% fuel losses during fabrication. Modifications of the current facility system and advanced casting techniques are underway to produce higher quality fuel slugs.

  16. Manufacture and Erection of SFR Components: Feedback from PFBR Experience

    International Nuclear Information System (INIS)

    Chellapandi, P.

    2013-01-01

    Unique Features of SFR Components: • Large diameter thin walled shell and slender structures calling for stringent tolerances posing challenges in manufacturing, handling and erection. • Single side welds are unavoidable at some difficult locations. • In-service inspection is difficult. • Residual stresses should be minimum calling for robust heat treatment strategy. • Minimum number of materials to be used from reliability point of view (but not preferred from economic considerations). • Mainly austenitic stainless steels calling for careful considerations for welding without significant weld repairs and distortions. • Reactor assembly components decide the project time schedule (large manufacturing, assembly and erection time). • Leak tightness is very important in view of resulting sodium leaks. • Limited experience on manufacturing and erection of components. • Design and manufacturing codes still evolvingPFBR Reactor Assembly – Major Lessons: • Grid plate Large number of sleeves, posing difficulty in assembly, hard facing of large diameter plates and heavy flange construction. • Roof slab Large box type structure with many penetrations – complicated manufacturing process, time consuming and difficulty to overcome lamellar tearing problems. • Inclined Fuel Transfer Machine Complex manufacturing processes leading to large time and extensive qualification tests. • Increase of number of primary pipes – essential for enhancing safety. • Integration of components manufactured by different industries took unduly long time

  17. Models used in the SFR1 SAR-08 and KBS-3H safety assessments for calculation of C-14 doses

    International Nuclear Information System (INIS)

    Avila, R.; Proehl, G.

    2008-03-01

    This report presents a set of simplified models for assessment of human exposures resulting from potential underground releases of C-14. These models were used in the SFR1 SAR08 and KBS-3H safety assessments. The proposed models can be used to assess continuous, as well as pulse-like C-14 releases, to various types of biosphere objects: forest ecosystems, agricultural lands, sea basins and lakes. It is also possible to make assessments of exposures resulting from the use of contaminated fresh waters, for example from an impacted well, for irrigation of vegetables. Models are also proposed for scenarios where lakes and sea basins are transformed into terrestrial objects due to land rise, filling of lakes and other natural or human induced processes. The exposure pathways considered in dose calculations with the models are: ingestion of contaminated food and water for both terrestrial and aquatic ecosystems, inhalation of contaminated air for terrestrial ecosystems. The exposure by external irradiation is not considered, as C-14 is a pure low energy beta emitter. The report provides an overview of the behaviour of C-14 in the environment, including an outline of the conceptual assumptions implicit in the proposed models. The proposed models are based on the so-called specific activity approach, which has been recommended by the UNSCEAR and the IAEA for assessment of doses resulting from C-14 releases to the environment from nuclear installations. The equations for estimation of the C-14 specific activities in environmental compartments have been derived from a combination of several realistic and conservative assumptions, which are documented and justified in the report. The models can be used in safety assessments of geological repositories of radioactive waste, to carry out cautious, but still not over conservative dose estimations, which can be compared with regulatory dose constrains. Comparative studies with the models indicate that the worse case situations

  18. Safety analysis for research reactors

    International Nuclear Information System (INIS)

    2008-01-01

    The aim of safety analysis for research reactors is to establish and confirm the design basis for items important to safety using appropriate analytical tools. The design, manufacture, construction and commissioning should be integrated with the safety analysis to ensure that the design intent has been incorporated into the as-built reactor. Safety analysis assesses the performance of the reactor against a broad range of operating conditions, postulated initiating events and other circumstances, in order to obtain a complete understanding of how the reactor is expected to perform in these situations. Safety analysis demonstrates that the reactor can be kept within the safety operating regimes established by the designer and approved by the regulatory body. This analysis can also be used as appropriate in the development of operating procedures, periodic testing and inspection programmes, proposals for modifications and experiments and emergency planning. The IAEA Safety Requirements publication on the Safety of Research Reactors states that the scope of safety analysis is required to include analysis of event sequences and evaluation of the consequences of the postulated initiating events and comparison of the results of the analysis with radiological acceptance criteria and design limits. This Safety Report elaborates on the requirements established in IAEA Safety Standards Series No. NS-R-4 on the Safety of Research Reactors, and the guidance given in IAEA Safety Series No. 35-G1, Safety Assessment of Research Reactors and Preparation of the Safety Analysis Report, providing detailed discussion and examples of related topics. Guidance is given in this report for carrying out safety analyses of research reactors, based on current international good practices. The report covers all the various steps required for a safety analysis; that is, selection of initiating events and acceptance criteria, rules and conventions, types of safety analysis, selection of

  19. Evaluation of Melt Behavior with initial Melt Velocity under SFR Severe Accidents

    Energy Technology Data Exchange (ETDEWEB)

    Heo, Hyo; Bang, In Cheol [UNIST, Ulsan (Korea, Republic of); Jerng, Dong Wook [Chung-Ang Univ, Seoul (Korea, Republic of)

    2015-10-15

    In the current Korean sodium-cooled fast reactor (SFR) program, early dispersion of the molten metallic fuel within a subchannel is suggested as one of the inherent safety strategies for the initiating phase of hypothetical core disruptive accident (HCDA). The safety strategy provides negative reactivity driven by the melt dispersal, so it could reduce the possibility of the recriticality event under a severe triple or more fault scenario for SFR. Since the behavior of the melt dispersion is unpredictable, it depends on the accident condition, particularly core region. While the voided coolant channel region is usually developed in the inner core, the unvoided coolant channel region is formed in the outer core. It is important to confirm the fuel dispersion with the core region, but there are not sufficient existing studies for them. From the existing studies, the coolant vapor pressure is considered as one of driving force to move the melt towards outside of the core. There is a complexity of the phenomena during intermixing of the melt with the coolant after the horizontal melt injections. It is too difficult to understand the several combined mechanisms related to the melt dispersion and the fragmentation. Thus, it could be worthwhile to study the horizontal melt injections at lower temperature as a preliminary study in order to identify the melt dispersion phenomena. For this reason, it is required to clarify whether the coolant vapor pressure is the driving force of the melt dispersion with the core region. The specific conditions to be well dispersed for the molten metallic fuel were discussed in the experiments with the simulant materials. The each melt behavior was compared to evaluate the melt dispersion under the coolant void condition and the boiling condition. As the results, the following results are remarked: 1. The upward melt dispersion did not occur for a given melt and coolant temperature in the nonboiling range. Over current range of conditions

  20. Evaluation of Melt Behavior with initial Melt Velocity under SFR Severe Accidents

    International Nuclear Information System (INIS)

    Heo, Hyo; Bang, In Cheol; Jerng, Dong Wook

    2015-01-01

    In the current Korean sodium-cooled fast reactor (SFR) program, early dispersion of the molten metallic fuel within a subchannel is suggested as one of the inherent safety strategies for the initiating phase of hypothetical core disruptive accident (HCDA). The safety strategy provides negative reactivity driven by the melt dispersal, so it could reduce the possibility of the recriticality event under a severe triple or more fault scenario for SFR. Since the behavior of the melt dispersion is unpredictable, it depends on the accident condition, particularly core region. While the voided coolant channel region is usually developed in the inner core, the unvoided coolant channel region is formed in the outer core. It is important to confirm the fuel dispersion with the core region, but there are not sufficient existing studies for them. From the existing studies, the coolant vapor pressure is considered as one of driving force to move the melt towards outside of the core. There is a complexity of the phenomena during intermixing of the melt with the coolant after the horizontal melt injections. It is too difficult to understand the several combined mechanisms related to the melt dispersion and the fragmentation. Thus, it could be worthwhile to study the horizontal melt injections at lower temperature as a preliminary study in order to identify the melt dispersion phenomena. For this reason, it is required to clarify whether the coolant vapor pressure is the driving force of the melt dispersion with the core region. The specific conditions to be well dispersed for the molten metallic fuel were discussed in the experiments with the simulant materials. The each melt behavior was compared to evaluate the melt dispersion under the coolant void condition and the boiling condition. As the results, the following results are remarked: 1. The upward melt dispersion did not occur for a given melt and coolant temperature in the nonboiling range. Over current range of conditions

  1. Preliminary structural integrity evaluations for the elevated temperature piping of the SFR IHTS against typical level a service events

    International Nuclear Information System (INIS)

    Park, Chang-Gyu; Kim, Jong-Bum; Lee, Jae-Han

    2009-01-01

    The SFR is adapting the IHTS(Intermediate Heat Transport System) to prevent the interaction of radioactive primary sodium and SG(Steam Generator) water. The IHTS hot leg piping connecting the IHX(Intermediate Heat eXchanger) to the SG of a 1200MWe pool-type SFR is an object component in this study. ASME Boiler and Pressure Vessel code Subsection NB provides rules for the design and analysis of the class 1 components. At an elevated temperature service, the ASME Subsection NH provides rules for the design and analysis of the Class 1 components but unfortunately, special rules for piping components are not provided until now. Therefore, the design and analysis of the IHTS hot leg piping shall comply with the design by analysis requirements of Subsection NH. The piping layout is proposed by considering the reactor component layout and reactor building space and the structural integrity is evaluated by considering two typical types of operating events in this study. Cycle type 1(CT-1) shows the refueling cycle event having a temperature history from a refueling temperature to a normal operating temperature via a hot standby temperature. Cycle type 2(CT-2) is a daily load follow operation. The structural integrity is evaluated by considering the enveloped CT-1 and CT-2 operating events per the ASME Subsection NH procedures. The SIE ASME-NH computer program, which has been developed to implement the ASME subsection NH rules, is used for the structural integrity evaluation by utilizing the finite element analysis results. (author)

  2. Development and Applicability Demonstration of a Remote Inspection Module for Inspection of Reactor Internals in an SFR

    International Nuclear Information System (INIS)

    Kim, Hoewoong; Joo, Youngsang; Park, Changgyu; Kim, Jongbum; Bae, Jinho

    2014-01-01

    Since liquid sodium is optically opaque, the ultrasonic inspection technique has been mainly employed for inspection of reactor internals in a Sodium-cooled Fast Reactor (SFR). Until now, two types of ultrasonic sensors have been mainly developed; immersion and waveguide sensors. An immersion sensor can provide a high-resolution image, but it may have problems in terms of reliability and life time because the sensor is exposed to high temperature during inspection. On the other hand, a waveguide sensor can maintain its performance during long-term inspection in high temperature because it installs an ultrasonic transducer in a cold region even though such a high-frequency ultrasonic wave cannot be used owing to the long propagation distance [4-6]. In this work, a remote inspection module employing four 10 m long waveguide sensors was newly developed and several performance tests were carried out in water to demonstrate the applicability of the developed remote inspection module to inspection of reactor internals in an SFR. In this work, a remote inspection module for inspection of reactor internals in an SFR was newly developed. The developed remote inspection module employs four 10 m long waveguide sensors for multiple inspection applications: a horizontal beam waveguide sensor for ranging inspection, two vertical beam waveguide sensors for viewing inspection and a 45 .deg. angle beam waveguide sensor for identification inspection. Several performance tests such as ranging, viewing and identification inspections were carried out for simulated nuclear fuel assembly specimens in water, and the applicability of the developed remote inspection module to inspection of reactor internals in an SFR was demonstrated

  3. Development and Applicability Demonstration of a Remote Inspection Module for Inspection of Reactor Internals in an SFR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hoewoong; Joo, Youngsang; Park, Changgyu; Kim, Jongbum [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Bae, Jinho [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    Since liquid sodium is optically opaque, the ultrasonic inspection technique has been mainly employed for inspection of reactor internals in a Sodium-cooled Fast Reactor (SFR). Until now, two types of ultrasonic sensors have been mainly developed; immersion and waveguide sensors. An immersion sensor can provide a high-resolution image, but it may have problems in terms of reliability and life time because the sensor is exposed to high temperature during inspection. On the other hand, a waveguide sensor can maintain its performance during long-term inspection in high temperature because it installs an ultrasonic transducer in a cold region even though such a high-frequency ultrasonic wave cannot be used owing to the long propagation distance [4-6]. In this work, a remote inspection module employing four 10 m long waveguide sensors was newly developed and several performance tests were carried out in water to demonstrate the applicability of the developed remote inspection module to inspection of reactor internals in an SFR. In this work, a remote inspection module for inspection of reactor internals in an SFR was newly developed. The developed remote inspection module employs four 10 m long waveguide sensors for multiple inspection applications: a horizontal beam waveguide sensor for ranging inspection, two vertical beam waveguide sensors for viewing inspection and a 45 .deg. angle beam waveguide sensor for identification inspection. Several performance tests such as ranging, viewing and identification inspections were carried out for simulated nuclear fuel assembly specimens in water, and the applicability of the developed remote inspection module to inspection of reactor internals in an SFR was demonstrated.

  4. Experimental validation of thermal design of top shield for a pool type SFR

    International Nuclear Information System (INIS)

    Aithal, Sriramachandra; Babu, V. Rajan; Balasubramaniyan, V.; Velusamy, K.; Chellapandi, P.

    2016-01-01

    Highlights: • Overall thermal design of top shield in a SFR is experimentally verified. • Air jet cooling is effective in ensuring the temperatures limits for top shield. • Convection patterns in narrow annulus are in line with published CFD results. • Wire mesh insulation ensures gradual thermal gradient at top portion of main vessel. • Under loss of cooling scenario, sufficient time is available for corrective action. - Abstract: An Integrated Top Shield Test Facility towards validation of thermal design of top shield for a pool type SFR has been conceived, constructed & commissioned. Detailed experiments were performed in this experimental facility having full-scale features. Steady state temperature distribution within the facility is measured for various heater plate temperatures in addition to simulating different operating states of the reactor. Following are the important observations (i) jet cooling system is effective in regulating the roof slab bottom plate temperature and thermal gradient across roof slab simulating normal operation of reactor, (ii) wire mesh insulation provided in roof slab-main vessel annulus is effective in obtaining gradual thermal gradient along main vessel top portion and inhibiting the setting up of cellular convection within annulus and (iii) cellular convection with four distinct convective cells sets in the annular gap between roof slab and small rotatable plug measuring ∼ϕ4 m in diameter & gap width varying from 16 mm to 30 mm. Repeatability of results is also ensured during all the above tests. The results presented in this paper is expected to provide reference data for validation of thermal hydraulic models in addition to serving as design validation of jet cooling system for pool type SFR.

  5. Evolution of Intrinsic Scatter in the SFR-Stellar Mass Correlation at 0.5 less than z Less Than 3

    Science.gov (United States)

    Kurczynski, Peter; Gawiser, Eric; Acquaviva, Viviana; Bell, Eric F.; Dekel, Avishai; De Mello, Duilia F.; Ferguson, Henry C.; Gardner, Jonathan P.; Grogin, Norman A.

    2016-01-01

    We present estimates of intrinsic scatter in the star formation rate (SFR)--stellar mass (M*) correlation in the redshift range 0.5 less than z less than 3.0 and in the mass range 10(exp 7) less than M* less than 10(exp 11) solar mass. We utilize photometry in the Hubble Ultradeep Field (HUDF12) and Ultraviolet Ultra Deep Field (UVUDF) campaigns and CANDELS/GOODS-S and estimate SFR, M* from broadband spectral energy distributions and the best-available redshifts. The maximum depth of the UDF photometry (F160W 29.9 AB, 5 sigma depth) probes the SFR--M* correlation down to M* approximately 10(exp 7) solar mass, a factor of 10-100 x lower in M* than previous studies, and comparable to dwarf galaxies in the local universe. We find the slope of the SFR-M* relationship to be near unity at all redshifts and the normalization to decrease with cosmic time. We find a moderate increase in intrinsic scatter with cosmic time from 0.2 to 0.4 dex across the epoch of peak cosmic star formation. None of our redshift bins show a statistically significant increase in intrinsic scatter approximately 100 Myr. Our results are consistent with a picture of gradual and self-similar assembly of galaxies across more than three orders of magnitude in stellar mass from as low as 10(exp 7) solar mass.

  6. Time Based Workload Analysis Method for Safety-Related Operator Actions in Safety Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yun Goo; Oh, Eung Se [Korea Hydro and Nuclear Power Co., Daejeon (Korea, Republic of)

    2016-05-15

    During the design basis event, the safety system performs safety functions to mitigate the event. The most of safety system is actuated by automatic system however, there are operator manual actions that are needed for the plant safety. These operator actions are classified as important human actions in human factors engineering design. The human factors engineering analysis and evaluation is needed for these important human actions to assure that operator successfully perform their tasks for plant safety and operational goals. The work load analysis is one of the required analysis for the important human actions.

  7. Time Based Workload Analysis Method for Safety-Related Operator Actions in Safety Analysis

    International Nuclear Information System (INIS)

    Kim, Yun Goo; Oh, Eung Se

    2016-01-01

    During the design basis event, the safety system performs safety functions to mitigate the event. The most of safety system is actuated by automatic system however, there are operator manual actions that are needed for the plant safety. These operator actions are classified as important human actions in human factors engineering design. The human factors engineering analysis and evaluation is needed for these important human actions to assure that operator successfully perform their tasks for plant safety and operational goals. The work load analysis is one of the required analysis for the important human actions.

  8. Operational experience from SFR - Final repository for low- and intermediate level waste in Sweden

    International Nuclear Information System (INIS)

    Skogsberg, Marie; Ingvarsson, Roger

    2006-01-01

    SFR, the Swedish Final Repository for Radioactive Waste, has been in operation since April 1988. It was designed for short lived LLW/ILW from the operation and maintenance of all Swedish Nuclear Power Plants. The first stage was constructed for 63 000 m 3 which was assumed to give a margin and flexibility for the preliminary operational period. Today this volume represents the whole prediction of operational waste. Until the end of 2005 SFR has received 30 930 m 3 waste. In average it has been 2-3 derivations per year at the repository. The most derivations happened in the years 1993-1995, and that was also the years when the repository received the most volume of waste. The most of the derivations those years was related to the waste packages. The dose rate to the personal has always been very low in the latest years the collective dose has been under 0,1 mmanSv/year. (author)

  9. Deterministic Safety Analysis for Nuclear Power Plants. Specific Safety Guide (Russian Edition)

    International Nuclear Information System (INIS)

    2014-01-01

    The objective of this Safety Guide is to provide harmonized guidance to designers, operators, regulators and providers of technical support on deterministic safety analysis for nuclear power plants. It provides information on the utilization of the results of such analysis for safety and reliability improvements. The Safety Guide addresses conservative, best estimate and uncertainty evaluation approaches to deterministic safety analysis and is applicable to current and future designs. Contents: 1. Introduction; 2. Grouping of initiating events and associated transients relating to plant states; 3. Deterministic safety analysis and acceptance criteria; 4. Conservative deterministic safety analysis; 5. Best estimate plus uncertainty analysis; 6. Verification and validation of computer codes; 7. Relation of deterministic safety analysis to engineering aspects of safety and probabilistic safety analysis; 8. Application of deterministic safety analysis; 9. Source term evaluation for operational states and accident conditions; References

  10. English translation of three documents relating to the SFR-1

    International Nuclear Information System (INIS)

    1988-01-01

    After approval from the National Institute of Radiation Protection, (the SSI) on April 26th, 1988 the Swedish Nuclear Fuel and Waste Management Company, the SKB, put the Final Repository for Radioactive Waste, the SFR-1 (Forsmark), into operation. This report contains English translations of the Operating Permission issued by SSI and the associated radiation protection instructions. Also included is a translation of chapter 4, the viewpoints and evaluations, of the Assessment Memorandum which was the background material for the Board of the SSI when deciding on the operational permission. (orig./HP)

  11. K Basin safety analysis

    International Nuclear Information System (INIS)

    Porten, D.R.; Crowe, R.D.

    1994-01-01

    The purpose of this accident safety analysis is to document in detail, analyses whose results were reported in summary form in the K Basins Safety Analysis Report WHC-SD-SNF-SAR-001. The safety analysis addressed the potential for release of radioactive and non-radioactive hazardous material located in the K Basins and their supporting facilities. The safety analysis covers the hazards associated with normal K Basin fuel storage and handling operations, fuel encapsulation, sludge encapsulation, and canister clean-up and disposal. After a review of the Criticality Safety Evaluation of the K Basin activities, the following postulated events were evaluated: Crane failure and casks dropped into loadout pit; Design basis earthquake; Hypothetical loss of basin water accident analysis; Combustion of uranium fuel following dryout; Crane failure and cask dropped onto floor of transfer area; Spent ion exchange shipment for burial; Hydrogen deflagration in ion exchange modules and filters; Release of Chlorine; Power availability and reliability; and Ashfall

  12. Safety design criteria for the next generation Sodium-cooled fast reactors based on lessons learned from the Fukushima NPS accident

    International Nuclear Information System (INIS)

    Sakai, Takaaki

    2012-01-01

    In this presentation, architecture of the safety design criteria as requirements for SFR system and the activities on safety research works to establish safety evaluation methods for the next generation SFRs are summarized with the basis on lessons learned from the Fukushima NPS accident. Nuclear safety is a grovel issue which should be achieved by the international cooperation. In respect of the development for the next generation reactor, it is necessary to build the harmonized safety criteria and evaluation methods to establish the next level of safety

  13. Coupled 3D-neutronics / thermal-hydraulics analysis of an unprotected loss-of-flow accident for a 3600 MWth SFR core

    International Nuclear Information System (INIS)

    Sun, K.; Chenu, A.; Mikityuk, K.; Krepel, J.; Chawla, R.

    2012-01-01

    The core behaviour of a large (3600 MWth) sodium-cooled fast reactor (SFR) is investigated in this paper with the use of a coupled TRACE/PARCS model. The SFR neutron spectrum is characterized by several performance advantages, but also leads to one dominating neutronics drawback - a positive sodium void reactivity. This implies a positive reactivity effect when sodium coolant is removed from the core. In order to evaluate such feedback in terms of the dynamics, a representative unprotected loss-of-flow (ULOF) transient, i.e. flow run-down without SCRAM in which sodium boiling occurs, is analyzed. Although analysis of a single transient cannot allow general conclusions to be drawn, it does allow better understanding of the underlying physics and can lead to proposals for improving the core response during such an accident. The starting point of this study is the reference core design considered in the framework of the Collaborative Project on the European Sodium Fast Reactor (CP-ESFR). To reduce the void effect, the core has been modified by introducing an upper sodium plenum (along with a boron layer) and by reducing the core height-to-diameter ratio. For the ULOF considered, a sharp increase in core power results in melting of the fuel in the case of the reference core. In the modified core, a large dryout leads to melting of the clad. It seems that, for the hypothetical event considered, fuel failure cannot be avoided with just improvement of the neutronics design; therefore, thermal-hydraulics optimization has been considered. An innovative assembly design is proposed to prevent sodium vapour blocking the fuel channel. This results in preventing a downward propagation of the sodium boiling to the core center, thus limiting it to the upper region. Such a void map introduces a negative coolant density reactivity feedback, which dominates the total reactivity change. As a result, the power level and the fuel temperature are effectively reduced, and a large dryout

  14. Coupled 3D-neutronics / thermal-hydraulics analysis of an unprotected loss-of-flow accident for a 3600 MWth SFR core

    Energy Technology Data Exchange (ETDEWEB)

    Sun, K. [Paul Scherrer Institut PSI, 5232 Villigen PSI (Switzerland); Ecole Polytechnique Federale de Lausanne EPFL, 1015 Lausanne (Switzerland); Chenu, A. [Ecole Polytechnique Federale de Lausanne EPFL, 1015 Lausanne (Switzerland); Mikityuk, K.; Krepel, J. [Paul Scherrer Institut PSI, 5232 Villigen PSI (Switzerland); Chawla, R. [Paul Scherrer Institut PSI, 5232 Villigen PSI (Switzerland); Ecole Polytechnique Federale de Lausanne EPFL, 1015 Lausanne (Switzerland)

    2012-07-01

    The core behaviour of a large (3600 MWth) sodium-cooled fast reactor (SFR) is investigated in this paper with the use of a coupled TRACE/PARCS model. The SFR neutron spectrum is characterized by several performance advantages, but also leads to one dominating neutronics drawback - a positive sodium void reactivity. This implies a positive reactivity effect when sodium coolant is removed from the core. In order to evaluate such feedback in terms of the dynamics, a representative unprotected loss-of-flow (ULOF) transient, i.e. flow run-down without SCRAM in which sodium boiling occurs, is analyzed. Although analysis of a single transient cannot allow general conclusions to be drawn, it does allow better understanding of the underlying physics and can lead to proposals for improving the core response during such an accident. The starting point of this study is the reference core design considered in the framework of the Collaborative Project on the European Sodium Fast Reactor (CP-ESFR). To reduce the void effect, the core has been modified by introducing an upper sodium plenum (along with a boron layer) and by reducing the core height-to-diameter ratio. For the ULOF considered, a sharp increase in core power results in melting of the fuel in the case of the reference core. In the modified core, a large dryout leads to melting of the clad. It seems that, for the hypothetical event considered, fuel failure cannot be avoided with just improvement of the neutronics design; therefore, thermal-hydraulics optimization has been considered. An innovative assembly design is proposed to prevent sodium vapour blocking the fuel channel. This results in preventing a downward propagation of the sodium boiling to the core center, thus limiting it to the upper region. Such a void map introduces a negative coolant density reactivity feedback, which dominates the total reactivity change. As a result, the power level and the fuel temperature are effectively reduced, and a large dryout

  15. Safety analysis for 'Fugen'

    International Nuclear Information System (INIS)

    1997-10-01

    The improvement of safety in nuclear power stations is an important proposition. Therefore also as to the safety evaluation, it is important to comprehensively and systematically execute it by referring to the operational experience and the new knowledge which is important for the safety throughout the period of use as well as before the construction and the start of operation of nuclear power stations. In this report, the results when the safety analysis for ''Fugen'' was carried out by referring to the newest technical knowledge are described. As the result, it was able to be confirmed that the safety of ''Fugen'' has been secured by the inherent safety and the facilities which were designed for securing the safety. The basic way of thinking on the safety analysis including the guidelines to be conformed to is mentioned. As to the abnormal transient change in operation and accidents, their definition, the events to be evaluated and the standards for judgement are reported. The matters which were taken in consideration at the time of the analysis are shown. The computation programs used for the analysis were REACT, HEATUP, LAYMON, FATRAC, SENHOR, LOTRAC, FLOOD and CONPOL. The analyses of the abnormal transient change in operation and accidents are reported on the causes, countermeasures, protective functions and results. (K.I.)

  16. A Review of Models for Dose Assessment Employed by SKB in the Renewed Safety Assessment for SFR 1

    Energy Technology Data Exchange (ETDEWEB)

    Shaw, George [Imperial College of Science Technology and Medicine (United Kingdom)

    2002-09-01

    This document provides a critical review, on behalf of SSI, of the models employed by the Swedish Nuclear Fuel and Waste Management Co (SKB) for dose assessment in the renewed safety assessment for the final repository for radioactive operational waste (SFR 1) in Forsmark, Sweden. The main objective of the review is to examine the models used by SKB for radiological dose assessment in a series of evolving biotopes in the vicinity of the Forsmark repository within a time frame beginning in 3000 AD and extending beyond 7500 AD. Five biosphere models (for coasts, lakes, agriculture, mires and wells) are described in Report TR-01-04. The principal consideration of the review is to determine whether these models are fit for the purpose of dose evaluation over the time frames involved and in the evolving sequence of biotopes specified. As well as providing general observations and comments on the modelling approach taken, six specific questions are addressed, as follows. Are the assumptions underlying the models justifiable? Are all reasonably foreseeable environmental processes considered? Has parameter uncertainty been sufficiently and reasonably addressed? Have sufficient models been used to address all reasonably foreseeable biotopes? Are the transitions between biotopes modelled adequately (specifically, are initial conditions for developing biotopes adequately specified by calculations for subsiding biotopes)? Have all critical radionuclides been identified? It is concluded that, in general, the assumptions underlying most of the models are justifiable. The exceptions are a) the rather simplistic approach taken in the Coastal Model and b) the lack of consideration of wild foods and age-dependence when calculating exposures of humans to radionuclides via dietary pathways. Most foreseeable processes appear to have been accounted for within the constraints of the models used, although it is recommended that attention be paid to future climate states when considering

  17. Reactor Safety Analysis

    International Nuclear Information System (INIS)

    Arien, B.

    2000-01-01

    The objective of SCK-CEN's programme on reactor safety is to develop expertise in probabilistic and deterministic reactor safety analysis. The research programme consists of two main activities, in particular the development of software for reliability analysis of large systems and participation in the international PHEBUS-FP programme for severe accidents. Main achievements in 1999 are reported

  18. System studies in PA: Development of process influence diagram (PID) for SFR-1 repository near-field + far-field

    International Nuclear Information System (INIS)

    Stenhouse, M.J.; Miller, W.M.; Chapman, N.A.

    2001-05-01

    Scenario development is a key component of the performance assessment (PA) process for radioactive waste disposal, the primary objective being to ensure that all relevant factors associated with the future evolution of the repository system are properly considered in PA. As part of scenario development, a list of features, events and processes (FEPs) are identified and assembled, representing the Process System, with interactions/influences between FEPs incorporated in a Process Influence Diagram (PID). This report documents the technical work conducted between 1997 and the end of 1999 under the Systems Studies Project. The overall objective of this project has been the construction of a PID for the SFR-1 repository (final repository for reactor waste), this PID being the first stage in the identification of scenarios to describe future evolution of this repository. The PIDs discussed in this report have been created using two software applications: existing commercial software (Business Modeller, Infotool AB. Stockholm, Sweden) and, more recently, a newly developed software tool SPARTA (Enviros QuantiSci, Henley, U.K.). Although the focus of this report is on the application of SPARTA to PID development, it is important to document the work carried out prior to SPARTA being available, in order to provide a complete record of the entire SFR-1 PID development effort as well as preserving the context of the multi-year project. Following a description of the different disposal sections of the SFR-1 and the various near-field barriers, the sequential development (i.e. near-field of Silo, BMA, BLA, BTF sections; far-field; integrated near-field + far-field) of the PID for SFR-1 repository system using Business Modeller is described. Owing to the complexity of the repository, in terms of number of both different disposal sections (Silo, BLA, BMA, BTF) and barriers associated with each section, the two-dimensional (2D) PID created for SFR-1 using Business Modeller is

  19. Design and Performance Evaluation of a Combined DHX unit for SFR Design Application

    International Nuclear Information System (INIS)

    Eoh, Jaehyuk; Kim, Dehee; Park, Chang-Gyu; Jeong, Ji-Young

    2015-01-01

    Based on a higher operating temperature with excellent thermal conductivity and larger thermal inertia of liquid sodium coolant, the SFR system has employed passive safety systems to ensure reliable decay heat removal (DHR) and consequential plant safety enhancement. Although a passive type DHR system has many advantages over an active one, designing a well coordinated passive system is usually more difficult than designing an effective active system. This is mainly because a cooling flow control is made directly by the system designer in an active system, while it is determined automatically by an intricate balance between the flow head loss and natural circulation head generation obtained from the density difference through the whole thermal flow system. To this end, securing a sufficient natural-circulation flow becomes one of the primary challenges for designing a reliable and successful Dh system in passive. In a current pool-type Sf design, an internal cooling flow path from the hot sodium pool to the cold pool is somewhat ambiguous owing to the split flow ratio formed in parallel paths between the intermediate heat exchangers (IHXs) and decay heat exchangers (DHXs), which results in a large uncertainty in the DHX shell-side flowrate and corresponding heat transfer to the DHR sodium loops. To improve passive the DHR performance, we proposed a new design concept with a simplified flow path from the hot pool to the cold pool through a unified flow path serially passing the DHX and IHX units. The present study aims at introducing the innovative design concept of the combined IHX-DHX unit and evaluating its design features in view of the heat transfer capability. From a comparison of the CHX performance designed by a one-dimensional approach with that made by a CFD analysis, it was quantitatively obtained that the difference in heat transfer rate is about 5.7%. It was also found that unexpected bypass flow in the shell-side CHX unit gave rise to a discrepancy

  20. Reliability analysis of PLC safety equipment

    Energy Technology Data Exchange (ETDEWEB)

    Yu, J.; Kim, J. Y. [Chungnam Nat. Univ., Daejeon (Korea, Republic of)

    2006-06-15

    FMEA analysis for Nuclear Safety Grade PLC, failure rate prediction for nuclear safety grade PLC, sensitivity analysis for components failure rate of nuclear safety grade PLC, unavailability analysis support for nuclear safety system.

  1. Reliability analysis of PLC safety equipment

    International Nuclear Information System (INIS)

    Yu, J.; Kim, J. Y.

    2006-06-01

    FMEA analysis for Nuclear Safety Grade PLC, failure rate prediction for nuclear safety grade PLC, sensitivity analysis for components failure rate of nuclear safety grade PLC, unavailability analysis support for nuclear safety system

  2. Safety Requirements / Design Criteria for SFR. Lessons Learned from the Fukushima Dai-ichi Accident

    International Nuclear Information System (INIS)

    Yllera, Javier

    2013-01-01

    After the Fukushima event (March 2011) the IAEA has started an action to review and revise, if necessary, all Safety Standards to take into consideration the lessons learned from the accident. The Safety Standards that need to be revised have been identified. A Prioritization Approach has been established: The first priority is to review safety guides applicable for NPPs and spent fuel storage with focus on the measures for the prevention and mitigation of severe accident due to external hazards - ● Regulatory framework, Safety assessment, Management system, Radiation protection and Emergency Preparedness and response; ● Sitting, Design, Operation of NPPs ● Decommissioning and Waste Management. Original sources for lessons learned: IAE fact Finding Mission, Japan´s report to the Ministerial Conference, INSAG Report, etc. Later, other lesson sources considered

  3. Reactor safety analysis

    International Nuclear Information System (INIS)

    Arien, B.

    1998-01-01

    Risk assessments of nuclear installations require accurate safety and reliability analyses to estimate the consequences of accidental events and their probability of occurrence. The objective of the work performed in this field at the Belgian Nuclear Research Centre SCK-CEN is to develop expertise in probabilistic and deterministic reactor safety analysis. The four main activities of the research project on reactor safety analysis are: (1) the development of software for the reliable analysis of large systems; (2) the development of an expert system for the aid to diagnosis; (3) the development and the application of a probabilistic reactor-dynamics method, and (4) to participate in the international PHEBUS-FP programme for severe accidents. Progress in research during 1997 is described

  4. Analysis of multiple failure accident scenarios for development of probabilistic safety assessment model for KALIMER-600

    International Nuclear Information System (INIS)

    Kim, T.W.; Suk, S.D.; Chang, W.P.; Kwon, Y.M.; Jeong, H.Y.; Lee, Y.B.; Ha, K.S.; Kim, S.J.

    2009-01-01

    A sodium-cooled fast reactor (SFR), KALIMER-600, is under development at KAERI. Its fuel is the metal fuel of U-TRU-Zr and it uses sodium as coolant. Its advantages are found in the aspects of an excellent uranium resource utilization, inherent safety features, and nonproliferation. The probabilistic safety assessment (PSA) will be one of the initiating subjects for designing it from the aspects of a risk informed design (RID) as well as a technology-neutral licensing (TNL). The core damage is defined as coolant voiding, fuel melting, or cladding damage. Accident scenarios which lead to the core damage should be identified for the development of a Level-1 PSA model. The SSC-K computer code is used to identify the conditions which lead to core damage. KALIMER-600 has passive safety features such as passive shutdown functions, passive pump coast-down features, and passive decay heat removal systems. It has inherent reactivity feedback effects such as Doppler, sodium void, core axial expansion, control rod axial expansion, core radial expansion, etc. The accidents which are analyzed are the multiple failure accidents such as an unprotected transient overpower, a loss of flow, and a loss of heat sink events with degraded safety systems or functions. The safety functions to be considered here are a reactor trip, inherent reactivity feedback features, the pump coast-down, and the passive decay heat removal. (author)

  5. Severe accident approach - final report. Evaluation of design measures for severe accident prevention and consequence mitigation

    International Nuclear Information System (INIS)

    Tentner, A.M.; Parma, E.; Wei, T.; Wigeland, R.

    2010-01-01

    An important goal of the US DOE reactor development program is to conceptualize advanced safety design features for a demonstration Sodium Fast Reactor (SFR). The treatment of severe accidents is one of the key safety issues in the design approach for advanced SFR systems. It is necessary to develop an in-depth understanding of the risk of severe accidents for the SFR so that appropriate risk management measures can be implemented early in the design process. This report presents the results of a review of the SFR features and phenomena that directly influence the sequence of events during a postulated severe accident. The report identifies the safety features used or proposed for various SFR designs in the US and worldwide for the prevention and/or mitigation of Core Disruptive Accidents (CDA). The report provides an overview of the current SFR safety approaches and the role of severe accidents. Mutual understanding of these design features and safety approaches is necessary for future collaborations between the US and its international partners as part of the GEN IV program. The report also reviews the basis for an integrated safety approach to severe accidents for the SFR that reflects the safety design knowledge gained in the US during the Advanced Liquid Metal Reactor (ALMR) and Integral Fast Reactor (IFR) programs. This approach relies on inherent reactor and plant safety performance characteristics to provide additional safety margins. The goal of this approach is to prevent development of severe accident conditions, even in the event of initiators with safety system failures previously recognized to lead directly to reactor damage.

  6. Severe accident approach - final report. Evaluation of design measures for severe accident prevention and consequence mitigation.

    Energy Technology Data Exchange (ETDEWEB)

    Tentner, A. M.; Parma, E.; Wei, T.; Wigeland, R.; Nuclear Engineering Division; SNL; INL

    2010-03-01

    An important goal of the US DOE reactor development program is to conceptualize advanced safety design features for a demonstration Sodium Fast Reactor (SFR). The treatment of severe accidents is one of the key safety issues in the design approach for advanced SFR systems. It is necessary to develop an in-depth understanding of the risk of severe accidents for the SFR so that appropriate risk management measures can be implemented early in the design process. This report presents the results of a review of the SFR features and phenomena that directly influence the sequence of events during a postulated severe accident. The report identifies the safety features used or proposed for various SFR designs in the US and worldwide for the prevention and/or mitigation of Core Disruptive Accidents (CDA). The report provides an overview of the current SFR safety approaches and the role of severe accidents. Mutual understanding of these design features and safety approaches is necessary for future collaborations between the US and its international partners as part of the GEN IV program. The report also reviews the basis for an integrated safety approach to severe accidents for the SFR that reflects the safety design knowledge gained in the US during the Advanced Liquid Metal Reactor (ALMR) and Integral Fast Reactor (IFR) programs. This approach relies on inherent reactor and plant safety performance characteristics to provide additional safety margins. The goal of this approach is to prevent development of severe accident conditions, even in the event of initiators with safety system failures previously recognized to lead directly to reactor damage.

  7. An overview of modeling methods for thermal mixing and stratification in large enclosures for reactor safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Haihua Zhao; Per F. Peterson

    2010-10-01

    Thermal mixing and stratification phenomena play major roles in the safety of reactor systems with large enclosures, such as containment safety in current fleet of LWRs, long-term passive containment cooling in Gen III+ plants including AP-1000 and ESBWR, the cold and hot pool mixing in pool type sodium cooled fast reactor systems (SFR), and reactor cavity cooling system behavior in high temperature gas cooled reactors (HTGR), etc. Depending on the fidelity requirement and computational resources, 0-D steady state models (heat transfer correlations), 0-D lumped parameter based transient models, 1-D physical-based coarse grain models, and 3-D CFD models are available. Current major system analysis codes either have no models or only 0-D models for thermal stratification and mixing, which can only give highly approximate results for simple cases. While 3-D CFD methods can be used to analyze simple configurations, these methods require very fine grid resolution to resolve thin substructures such as jets and wall boundaries. Due to prohibitive computational expenses for long transients in very large volumes, 3-D CFD simulations remain impractical for system analyses. For mixing in stably stratified large enclosures, UC Berkeley developed 1-D models basing on Zuber’s hierarchical two-tiered scaling analysis (HTTSA) method where the ambient fluid volume is represented by 1-D transient partial differential equations and substructures such as free or wall jets are modeled with 1-D integral models. This allows very large reductions in computational effort compared to 3-D CFD modeling. This paper will present an overview on important thermal mixing and stratification phenomena in large enclosures for different reactors, major modeling methods and their advantages and limits, potential paths to improve simulation capability and reduce analysis uncertainty in this area for advanced reactor system analysis tools.

  8. Overview of Generation IV (Gen IV) Reactor Designs - Safety and Radiological Protection Considerations

    International Nuclear Information System (INIS)

    Baudrand, Olivier; Blanc, Daniel; Ivanov, Evgeny; Bonneville, Herve; Clement, Bernard; Kissane, Martin; Meignen, Renaud; Monhardt, Daniel; Nicaise, Gregory; Bourgois, Thierry; Bruna, Giovanni; Hache, Georges; Repussard, Jacques

    2012-01-01

    The purpose of this document is to provide an updated overview of specific safety and radiological protection issues for all the reactor concepts adopted by the GIF (Generation IV International Forum), independent of their advantages or disadvantages in terms of resource optimization or long-lived-waste reduction. In particular, this new document attempts to bring out the advantages and disadvantages of each concept in terms of safety, taking into account the Western European Nuclear Regulators' Association (WENRA) statement concerning safety objectives for new nuclear power plants. Using an identical framework for each reactor concept (sodium-cooled fast reactors or SFR, high / very-high temperature helium-cooled reactors of V/HTR, gas-cooled fast reactors or GFR, lead-or lead / bismuth-cooled fast reactors or LFR, molten salt reactors or MSR, and supercritical-water-cooled reactors or SCWR), this summary report provides some general conclusions regarding their safety and radiological protection issues, inspired by WENRA's safety objectives and on the basis of available information. Initial lessons drawn from the events at the Fukushima-Daiichi nuclear power plant in March 2011 have also been taken into account in IRSN's analysis of each reactor concept

  9. Uncertainties in estimating the {sup 90}Sr and actinides inventory in SFR 1; Osaekerheter vid uppskattning av Sr-90 och aktinidinventariet i SFR 1

    Energy Technology Data Exchange (ETDEWEB)

    Ingemansson, Tor [ALARA Engineering AB, Skultuna (Sweden)

    2000-04-01

    SFR-1 is a facility for disposal of low and intermediate level radioactive waste. The uncertainty in estimation of the activity accumulated in different cleaning filters, originating in the Swedish BWR-, PWR-reactors and CLAB - the Central interim storage facility for spent nuclear fuel - has been analyzed to be 10 - 14%, depending on the methods used for measuring the activity at the power plants. Other waste or scrap contribute with approx. 1.5% of the total amount of actinides and {sup 90}Sr. The uncertainty in this fraction is about 20%. The uncertainties are surprisingly small, and explain the good agreement between estimates made with different methods.

  10. Visualization Study of Melt Dispersion Behavior for SFR with a Metallic Fuel under Severe Accidents

    Energy Technology Data Exchange (ETDEWEB)

    Heo, Hyo Heo; Park, Seong Dae; Bang, In Cheol [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of); Jerng, Dong Wook [Jungang Univ., Seoul (Korea, Republic of)

    2015-05-15

    The safety strategy provides negative reactivity driven by the melt dispersal, so it could reduce the possibility of the recriticality event under a severe triple or more fault scenario for SFR. Since the behavior of the melt dispersion is unpredictable, it depends on the accident condition, particularly core region. While the voided coolant channel region is usually developed in the inner core, the unvoided coolant channel region is formed in the outer core. It is important to confirm the fuel dispersion with the core region, but there are not sufficient existing studies for them. From the existing studies, the coolant vapor pressure is considered as one of driving force to move the melt towards outside of the core. There is a complexity of the phenomena during intermixing of the melt with the coolant after the horizontal melt injections. It is too difficult to understand the several combined mechanisms related to the melt dispersion and the fragmentation. The specific conditions to be well dispersed for the molten metallic fuel were discussed in the experiments with the simulant materials. The each melt behavior was compared to evaluate the melt dispersion under the coolant void condition and the boiling condition.

  11. Periodic safety review of the HTR-10 safety analysis

    International Nuclear Information System (INIS)

    Chen Fubing; Zheng Yanhua; Shi Lei; Li Fu

    2015-01-01

    Designed by the Institute of Nuclear and New Energy Technology (INET) of Tsinghua University, the 10 MW High Temperature Gas-cooled Reactor-Test Module (HTR-10) is the first modular High Temperature Gas-cooled Reactor (HTGR) in China. According to the nuclear safety regulations of China, the periodic safety review (PSR) of the HTR-10 was initiated by INET after approved by the National Nuclear Safety Administration (NNSA) of China. Safety analysis of the HTR-10 is one of the key safety factors of the PSR. In this paper, the main contents in the review of safety analysis are summarized; meanwhile, the internal evaluation on the review results is presented by INET. (authors)

  12. Safety analysis of nuclear power plants

    International Nuclear Information System (INIS)

    Selvatici, E.

    1981-01-01

    A study about the safety analysis of nuclear power plant, giving emphasis to how and why to do is presented. The utilization of the safety analysis aiming to perform the licensing requirements is discussed, and an example of the Angra 2 and 3 safety analysis is shown. Some presented tendency of the safety analysis are presented and examples are shown.(E.G.) [pt

  13. Investigation of the Stress Intensity Limits of ASME Section III Div.5 for Structure Design Criteria of SFR Fuel Assembly

    Energy Technology Data Exchange (ETDEWEB)

    Choo, Jin-Yup; Kim, Hyung-Kyu; Cheon, Jin-Sik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    These affect the mechanical design of the fuel assembly components. And thus, appropriate structural design criteria should also be chosen to incorporate the specific design conditions of the SFR fuel assemblies. Among them, the temperature is one of the most crucial conditions to be concerned because the sodium coolant temperature is normally more than 500ºC which is much higher than that of the LWR (< 350ºC). This implies that a thermal creep should be significantly considered in the SFR fuel assembly mechanical design. In addition to the high temperature condition, an irradiation swelling is also an important behavior that the SFR fuel assembly material should accommodate. To incorporate the temperature and irradiation impacts, the material of the fuel assembly components is presently determined to be made of HT-9, the ferriticmartensitic steel. In this paper, the ASME Sec. III Div. 5 (referred to as ‘Div. 5’ hereinafter), which was developed for a ‘high temperature reactor’, is considered as one of the structural design criteria for the mechanical design of SFR fuel assemblies. In this paper, the stress intensity limits, S{sub m} and S{sub t} of HT-9 were built for the structural criteria of an SFR fuel assembly. S{sub m} is obtained from the ultimate strength. As for S{sub t}, it is more complicated because of its dependency of time duration in addition to temperature. Following the definition of S{sub mt}, the method in the ASME Sec. III Div. 1, Subsec. NH was consulted. We found that the Sm is adopted as S{sub mt} under the temperature about 470ºC which is relatively low temperature range and over 470ºC with relatively short time duration as 1000 hours. And the S{sub t} is adopted as Smt at over 470ºC and long time duration over 34800 hours, and over 520ºC and 10{sup 4} hours too. And at over 570ºC and 1000 hours, and at over 630ºC and 100 hours, S{sub t} is also adopted for S{sub mt}.

  14. 10 CFR 70.62 - Safety program and integrated safety analysis.

    Science.gov (United States)

    2010-01-01

    ...; (iv) Potential accident sequences caused by process deviations or other events internal to the... have experience in nuclear criticality safety, radiation safety, fire safety, and chemical process... this safety program; namely, process safety information, integrated safety analysis, and management...

  15. The practical implementation of integrated safety management for nuclear safety analysis and fire hazards analysis documentation

    International Nuclear Information System (INIS)

    COLLOPY, M.T.

    1999-01-01

    In 1995 Mr. Joseph DiNunno of the Defense Nuclear Facilities Safety Board issued an approach to describe the concept of an integrated safety management program which incorporates hazard and safety analysis to address a multitude of hazards affecting the public, worker, property, and the environment. Since then the U S . Department of Energy (DOE) has adopted a policy to systematically integrate safety into management and work practices at all levels so that missions can be completed while protecting the public, worker, and the environment. While the DOE and its contractors possessed a variety of processes for analyzing fire hazards at a facility, activity, and job; the outcome and assumptions of these processes have not always been consistent for similar types of hazards within the safety analysis and the fire hazard analysis. Although the safety analysis and the fire hazard analysis are driven by different DOE Orders and requirements, these analyses should not be entirely independent and their preparation should be integrated to ensure consistency of assumptions, consequences, design considerations, and other controls. Under the DOE policy to implement an integrated safety management system, identification of hazards must be evaluated and agreed upon to ensure that the public. the workers. and the environment are protected from adverse consequences. The DOE program and contractor management need a uniform, up-to-date reference with which to plan. budget, and manage nuclear programs. It is crucial that DOE understand the hazards and risks necessarily to authorize the work needed to be performed. If integrated safety management is not incorporated into the preparation of the safety analysis and the fire hazard analysis, inconsistencies between assumptions, consequences, design considerations, and controls may occur that affect safety. Furthermore, confusion created by inconsistencies may occur in the DOE process to grant authorization of the work. In accordance with

  16. A Preliminary Design Study of Ultra-Long-Life SFR Cores having Heterogeneous Fuel Assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Jung, GeonHee; You, WuSeung; Hong, Ser Gi [Kyung Hee University, Yongin (Korea, Republic of)

    2016-10-15

    The PWR and CANDU reactors have provided electricity for several decades in our country but they have produced lots of spent fuels and so the safe and efficient disposal of these spent fuels is one of the main issues in nuclear industry. This type ultra-long-life cores are quite efficient in terms of the amount of spent fuel generation per electricity production and they can be used as an interim storage for PWR or CANDU spent fuel over several tens of years if they use the PWR or CANDU spent fuel as the initial fuel. Typically, the previous works have considered radially homogeneous fuel assemblies in which only blanket or driver fuel rods are employed and they considered axially or radially heterogeneous core configurations with the radially homogeneous fuel assemblies. These core configurations result in the propagation of the power distribution which can lead to the significant temperature changes for each fuel assembly over the time. In this work, the radially heterogeneous fuel assemblies are employed in new ultra-long-life SFR (Sodium-cooled Fast Reactor) cores to minimize the propagation of power distribution by allowing the power propagation in the fuel assemblies. In this work, new small ultra-long life SFR cores were designed with heterogeneous fuel assemblies having both blanket and driver fuel rods to minimize the propagation of power distribution over the core by allowing power propagation from driver rods to blanket rods in fuel assemblies. In particular, high fidelity depletion calculation coupled with heterogeneous Monte Carlo neutron transport calculation was performed to assess the neutronic feasibility of the ultralong life cores. The results of the analysis showed that the candidate core has the cycle length of 77 EFPYs, a small burnup reactivity swing of 1590 pcm and acceptably small SVRs both at BOC and EOC.

  17. Development of the Level 1 PSA Model for PGSFR Regulatory

    International Nuclear Information System (INIS)

    Na, Hyun Ju; Lee, Yong Suk; Shin, Andong; Suh, Nam Duk

    2014-01-01

    SFR (sodium-cooled fast reactor) is Gen-IV nuclear energy system, which is designed for stability, sustainability and proliferation resistance. KALIMER-600 and PGSFR (Prototype Gen-IV SFR) are under development in Korea with enhanced passive safety concepts, e.g. passive reactor shutdown, passive residual heat removal, and etc. Risk analysis from a regulatory perspective is necessary for regulatory body to support the safety and licensing review of SFR. Safety issues should be identified in the early design phase in order to prevent the unexpected cost increase and the delay of PGSFR licensing schedule. In this respect, the preliminary PSA Model of KALIMER-600 had been developed for regulatory. In this study, the development of PSA Level 1 Model is presented. The important impact factors in the risk analysis for the PGSFR, such as Core Damage Frequency (CDF), have been identified and the related safety insights have been derived. The PSA level 1 model for PGSFR regulatory is developed and the risk analysis is conducted. Regarding CDF, LOISF frequency, uncertainty parameter for passive system CCF, loss of 125V DC control center bus and damper CCF are identified as the important factors. Sensitivity analyses show that the CDF would be differentiated (lowered) according to their values

  18. IAEA Review for Gap Analysis of Safety Analysis Capability

    International Nuclear Information System (INIS)

    Basic, Ivica; Kim, Manwoong; Huges, Peter; Lim, B-K; D'Auria, Francesco; Louis, Vidard Michael

    2014-01-01

    The IAEA Asian Nuclear Safety Network (ANSN) was launched in 2002 in the framework of the Extra Budgetary Programme (EBP) on the Safety of Nuclear Installations in the South East Asia, Pacific and Far East Countries. The main objective is to strengthen and expand human and advanced Information Technology (IT) network to pool, analyse and share nuclear safety knowledge and practical experience for peaceful uses in this region. Under the ANSN framework, a technical group on Safety Analysis (SATG) was established in 2004 aimed to providing a forum for the exchange of experience in the following areas of safety analysis: · To provide a forum for an exchange of experience in the area of safety analysis, · To maintain and improve the knowledge on safety analysis method, · To enhance the utilization of computer codes, · To pool and analyse the issues related with safety analysis of research reactor, and · To facilitate mutual interested on safety analysis among member countries. A sustainable and successful nuclear energy programme requires a strong technical infrastructure, including a workforce made up of highly specialized and well-educated professionals. A significant portion of this technical capacity must be dedicated to safety- especially to safety analysis- as only then can it serve as the basis for making the right decisions during the planning, licensing, construction and operation of new nuclear facilities. In this regard, the IAEA has provided ANSN member countries with comprehensive training opportunities for capacity building in safety analysis. Nevertheless, the SATG recognizes that it is difficult to achieve harmonization in this area among all member countries because of their different competency levels. Therefore, it is necessary to quickly identify the most obvious gaps in safety analysis capability and then to use existing resources to begin to fill those gaps. The goal of this Expert Mission (EM) for gap finding service is to facilitate

  19. Basic Design of Experimental Facility for Measuring Pressure Drop of IHX in a SFR

    Energy Technology Data Exchange (ETDEWEB)

    Ko, Yung-Joo; Eoh, Jae-Hyuk; Kim, Hyungmo; Lee, Dong-Won; Jeong, Ji-Young; Lee, Hyeong-Yeon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Dong Eok [Kyungpook National Univ., Daegu (Korea, Republic of)

    2015-05-15

    The conceptual design of the Prototype gen-IV SFR (PGSFR) with a 150 MWe capacity was commenced in 2012 through the national long-term R and D program by KAERI. Then, PGSFR is now being designed with the defense in depth concept with active, passive and inherent safety features to acquire design approval for PGSFR from the Korean regulatory authority by 2020. PGSFR is a sodium-cooled pool-type fast reactor with all primary components including the primary heat transport system (PHTS) pumps and IHXs are located inside a sodium pool. The heat produced due to fission in the core is transported by primary sodium to secondary sodium in a sodium to sodium intermediate heat exchanger (IHX), which in turn is transferred to water in a steam generator (SG). Basic design of the IHX flow characteristic test facility, WEIPA was conducted based on the three-level scaling methodology in order to preserve the flow characteristics of the IHX in PGSFR. This test facility is intended to measure a high precision pressure drop at the shell-side of the IHX. This paper describes the aspects of the current design features of the IHX in PGSFR, scaling and basic design features of the facility.

  20. Reactor Safety Analysis

    International Nuclear Information System (INIS)

    Arien, B.

    1998-01-01

    The objective of SCK-CEN's programme on reactor safety is to develop expertise in probabilistic and deterministic reactor safety analysis. The research programme consists of four main activities, in particular the development of software for reliability analysis of large systems and participation in the international PHEBUS-FP programme for severe accidents, the development of an expert system for the aid to diagnosis; the development and application of a probabilistic reactor dynamics method. Main achievements in 1999 are reported

  1. Severe accidents and ESFR design issues

    International Nuclear Information System (INIS)

    Rineiski, A.

    2013-01-01

    Current SFR studies in Germany: ⇒ In support of European SFR studies, mainly on safety and safety-related (design optimization) issues; ⇒ ADS and SFR as main options for spent fuel management in studies on the possibility of P&T; ⇒ ESFR-type designs studied recently; ⇒ ASTRID-type designs to be studied in the future; ⇒ Particular area: modeling of severe accidents with SAS4A/SAS-SFR and SIMMER codes

  2. Conceptual Design for BOP of the Sodium-Cooled Fast Reactor

    International Nuclear Information System (INIS)

    Yoo, Tae Geun; Kim, Seong O; Kim, Eui Kwang; Seong, Seung Hwan

    2010-01-01

    The heavy dependence on nuclear power eventually raise the issues of an efficient utilization of uranium resources, which Korea presently imports from abroad, end of a spent fuel storage. From the viewpoint that sodium-cooled fast Reactors (SFR s ) have the potential of an enhanced safety by utilizing inherent safety characteristics, trans-uranics (TRU) reduction and resolving the spent fuel storage problems through a proliferation-resistant actinide recycling. SFR s are sure to be most promising nuclear power operation. The Korea Atomic Energy Research Institute (KAERI) has been developing SFR design technologies since 1997. And nowadays, the preliminary heat balance of the demonstration SFR is calculated. However, in order to verify design condition of the NSSS, it is necessary to set the heat balance and the conceptual design for BOP of the SFR as a part of the SFR design technique development business. Moreover, in order to confirm whether the heat balance can actually appropriate via the turbine characteristic, it is required to carry out the performance analysis of the turbine cycle. For that, the main purposes of this study are; 1) to derivate the conceptual design for BOP, 2) to analyze the performance of the turbine cycle, 3) to derivate the main consideration for BOP design

  3. Minutes of the 2. Meeting of the WPRS / EGRPANS / Sodium Fast Reactor Task Force (SFR)

    International Nuclear Information System (INIS)

    Ivanov, Evgeny; Kereszturi, Andras; Pataki, I.; Tota, A.; Vertes, P.; Kim, Taek K.; Taiwo, T.A.; Kugo, Teruhiko; Lee, Yi Kang; Messaoudi, Nadia; Michel-Sendis, Franco; ); Pascal, Vincent; Buiron, Laurent; Varaine, Frederic; Ponomarev, Alexander

    2012-01-01

    Five organizations (SCK/CEN, KIT, KFKI, CEA, ANL) participated in the Sodium-cooled fast reactor (SFR) Benchmark calculations and all results were collected and compiled by CEA and ANL. The compiled results of the large size cores and medium size cores were presented by V. Pascal (CEA) and T. K. Kim (ANL), respectively. Separately, A. Kereszturi presented his recently updated results. It was observed that there is wide variation in core multiplication factor, kinetics parameters, and reactivity feedback coefficients. In particular, compared to the CEA results, ANL calculated smaller k-eff, Doppler constant, but higher sodium void worth and control rod worth. The core modeling issue (heterogeneous vs. homogeneous) and solution method (diffusion vs. transport) were identified as the potential reasons of these discrepancies, including the minor impacts from the depletion chains and lumped fission product modeling. All participants agreed that additional investigation was needed to identify the reasons of these discrepancies. In addition, V. Pascal presented the informative notes of the reactivity feedback calculations methodology proposed by CEA. This document brings together the 5 presentations (slides) given at this meeting: 1 - SFR Task Force : Core behavior during transient as a function of power size and fuel nature (L. Buiron, V. Pascal, F. Varaine); 2 - Sodium Fast Reactor core Feedback and Transient response (SFRFT) Expert Group: preliminary benchmark results for large cores (L. Buiron, V. Pascal, F. Varaine); 3 - Numerical Benchmark Results for 1000 MWth Sodium-cooled Fast Reactor (T.K. Kim and T.A. Taiwo); 4 - Preliminary results of the WPRS Sodium-Cooled Fast Reactor Benchmark problems (A. Kereszturi, I. Pataki, A. Tota, P. Vertes); 5 - SFR Task Force : proposal for Feedback coefficients estimation methodology (L. Buiron, V.Pascal, F. Varaine)

  4. System studies in PA: Development of process influence diagram (PID) for SFR-1 repository near-field + far-field

    Energy Technology Data Exchange (ETDEWEB)

    Stenhouse, M.J. [Monitor Scientific, LLC, Denver, CO (United States); Miller, W.M.; Chapman, N.A. [QuantiSci Ltd., Melton Mowbray (United Kingdom)

    2001-05-01

    Scenario development is a key component of the performance assessment (PA) process for radioactive waste disposal, the primary objective being to ensure that all relevant factors associated with the future evolution of the repository system are properly considered in PA. As part of scenario development, a list of features, events and processes (FEPs) are identified and assembled, representing the Process System, with interactions/influences between FEPs incorporated in a Process Influence Diagram (PID). This report documents the technical work conducted between 1997 and the end of 1999 under the Systems Studies Project. The overall objective of this project has been the construction of a PID for the SFR-1 repository (final repository for reactor waste), this PID being the first stage in the identification of scenarios to describe future evolution of this repository. The PIDs discussed in this report have been created using two software applications: existing commercial software (Business Modeller, Infotool AB. Stockholm, Sweden) and, more recently, a newly developed software tool SPARTA (Enviros QuantiSci, Henley, U.K.). Although the focus of this report is on the application of SPARTA to PID development, it is important to document the work carried out prior to SPARTA being available, in order to provide a complete record of the entire SFR-1 PID development effort as well as preserving the context of the multi-year project. Following a description of the different disposal sections of the SFR-1 and the various near-field barriers, the sequential development (i.e. near-field of Silo, BMA, BLA, BTF sections; far-field; integrated near-field + far-field) of the PID for SFR-1 repository system using Business Modeller is described. Owing to the complexity of the repository, in terms of number of both different disposal sections (Silo, BLA, BMA, BTF) and barriers associated with each section, the two-dimensional (2D) PID created for SFR-1 using Business Modeller is

  5. Systems engineered health and safety criteria for safety analysis reports

    International Nuclear Information System (INIS)

    Beitel, G.A.; Morcos, N.

    1993-01-01

    The world of safety analysis is filled with ambiguous words: codes and standards, consequences and risks, hazard and accident, and health and safety. These words have been subject to disparate interpretations by safety analysis report (SAR) writers, readers, and users. open-quotes Principal health and safety criteriaclose quotes has been one of the most frequently misused phrases; rarely is it used consistently or effectively. This paper offers an easily understood definition for open-quotes principal health and safety criteriaclose quotes and uses systems engineering to convert an otherwise mysterious topic into the primary means of producing an integrated SAR. This paper is based on SARs being written for environmental restoration and waste management activities for the U.S. Department of Energy (DOE). Requirements for these SARs are prescribed in DOE Order 5480-23, open-quotes Nuclear Safety Analysis Reports.close quotes

  6. Study of the occurrence of organic matter, metals and chemicals in the SFR

    International Nuclear Information System (INIS)

    Sundqvist, J.O.

    2001-03-01

    Low- and intermediate level operational waste from the Swedish nuclear power plants, and the Studsvik facility, is currently placed in a repository, termed SFR-l (final repository for radioactive operational waste) near the Forsmark power plant. Two important components in the waste, which can affect the function of the repository, are organic materials, e.g. cloth and paper, and metals (scrap). The release of radionuclides from the repository may be affected by chemical reactions that involve both organic materials and metals. After sealing the repository, the conditions can be such that complexing agents (e.g. isosaccarinic acid) may form when organic materials degrade. These agents typically increase the mobility of radionuclides. Formation of gas, mainly due to metal corrosion, may affect the barrier system, surrounding the waste, such that the release of radionuclides is enhanced. SKB makes an annual report with a compilation of the waste that has been placed in SFR-l . The compilation contains both the amount of waste placed in the repository during the last year and a compilation of the waste that have been placed since the stall of SFR. Moreover, SKB provides a prognosis of the future situation in SFR-1 every third year. SKI (the Swedish Nuclear Power Inspectorate), is responsible for reviewing this reporting. This study was initiated with the purpose of evaluating the uncertainties in SKB's estimates of the amounts of organic matter, metals and chemicals in the waste in SFR- I. The estimates of the quantities of e.g. cellulose and metals in the waste are based on a method which is utilising what is called normal-containers. The waste is classified into certain waste categories. For each waste category there is a specified, presumed composition, named normal-container. The results of this study suggest that the documentation provided by SKB is lacking in some respects. There are for instance examples of incomplete notification of waste and container types

  7. Ignalina Safety Analysis Group

    International Nuclear Information System (INIS)

    Ushpuras, E.

    1995-01-01

    The article describes the fields of activities of Ignalina NPP Safety Analysis Group (ISAG) in the Lithuanian Energy Institute and overview the main achievements gained since the group establishment in 1992. The group is working under the following guidelines: in-depth analysis of the fundamental physical processes of RBMK-1500 reactors; collection, systematization and verification of the design and operational data; simulation and analysis of potential accident consequences; analysis of thermohydraulic and neutronic characteristics of the plant; provision of technical and scientific consultations to VATESI, Governmental authorities, and also international institutions, participating in various projects aiming at Ignalina NPP safety enhancement. The ISAG is performing broad scientific co-operation programs with both Eastern and Western scientific groups, supplying engineering assistance for Ignalina NPP. ISAG is also participating in the joint Lithuanian - Swedish - Russian project - Barselina, the first Probabilistic Safety Assessment (PSA) study of Ignalina NPP. The work is underway together with Maryland University (USA) for assessment of the accident confinement system for a range of breaks in the primary circuit. At present the ISAG personnel is also involved in the project under the grant from the Nuclear Safety Account, administered by the European Bank for reconstruction and development for the preparation and review of an in-depth safety assessment of the Ignalina plant

  8. Preliminary safety analysis methodology for the SMART

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Kyoo Hwan; Chung, Y. J.; Kim, H. C.; Sim, S. K.; Lee, W. J.; Chung, B. D.; Song, J. H. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-03-01

    This technical report was prepared for a preliminary safety analysis methodology of the 330MWt SMART (System-integrated Modular Advanced ReacTor) which has been developed by Korea Atomic Energy Research Institute (KAERI) and funded by the Ministry of Science and Technology (MOST) since July 1996. This preliminary safety analysis methodology has been used to identify an envelope for the safety of the SMART conceptual design. As the SMART design evolves, further validated final safety analysis methodology will be developed. Current licensing safety analysis methodology of the Westinghouse and KSNPP PWRs operating and under development in Korea as well as the Russian licensing safety analysis methodology for the integral reactors have been reviewed and compared to develop the preliminary SMART safety analysis methodology. SMART design characteristics and safety systems have been reviewed against licensing practices of the PWRs operating or KNGR (Korean Next Generation Reactor) under construction in Korea. Detailed safety analysis methodology has been developed for the potential SMART limiting events of main steam line break, main feedwater pipe break, loss of reactor coolant flow, CEA withdrawal, primary to secondary pipe break and the small break loss of coolant accident. SMART preliminary safety analysis methodology will be further developed and validated in parallel with the safety analysis codes as the SMART design further evolves. Validated safety analysis methodology will be submitted to MOST as a Topical Report for a review of the SMART licensing safety analysis methodology. Thus, it is recommended for the nuclear regulatory authority to establish regulatory guides and criteria for the integral reactor. 22 refs., 18 figs., 16 tabs. (Author)

  9. Transient analysis for resolving safety issues

    International Nuclear Information System (INIS)

    Chao, J.; Layman, W.

    1987-01-01

    The Nuclear Safety Analysis Center (NSAC) has a Generic Safety Analysis Program to help resolve high priority generic safety issues. This paper describes several high priority safety issues considered at NSAC and how they were resolved by transient analysis using thermal hydraulics and neutronics codes. These issues are pressurized thermal shock (PTS), anticipated transients without scram (ATWS), steam generator tube rupture (SGTR), and reactivity transients in light of the Chernobyl accident

  10. Status of the design and safety project for the sodium-cooled fast reactor as a generation IV nuclear energy system

    International Nuclear Information System (INIS)

    Niwa, Hajime; Fiorini, Gian-Luigi; Sim, Yoon-Sub; Lennox, Tom; Cahalan, James E.

    2005-01-01

    The Design and Safety Project Management Board (DSPMB) was established under the Sodium Cooled Fast Reactor (SFR) System Steering Committee (SSC) in the Generation IV international Forum. The DSPMB will promote collaborative R and D activities on reactor core design, and safety assessment for candidate systems, and also integrate these results together with those from other PMBs such as advanced fuel and component to a whole fast reactor system in order to develop high performance systems that will satisfy the goals of Generation IV nuclear energy systems. The DSPMB has formulated the present R and D schedules for this purpose. Two SFR concepts were proposed: a loop-type system with primarily a MOX fuel core and a pool-type system with a metal fuel core. Study of innovative systems and their evaluation will also be included. The safety project will cover both the safety assessment of the design and the preparation of the methods/tools to be used for the assessment. After a rather short viability phase, the project will move to the performance phase for development of performance data and design optimization of conceptual designs. This paper describes the schedules, work packages and tasks for the collaborative studies of the member countries. (author)

  11. Pre-Licensing Evaluation of Legacy SFR Metallic Fuel Data

    Energy Technology Data Exchange (ETDEWEB)

    Yacout, A. M. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Billone, M. C. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division

    2016-09-16

    The US sodium cooled fast reactor (SFR) metallic fuel performance data that are of interest to advanced fast reactors applications, can be attributed mostly to the Integral Fast Reactor (IFR) program between 1984 and 1994. Metallic fuel data collected prior to the IFR program were associated with types of fuel that are not of interest to future advanced reactors deployment (e.g., previous U-Fissium alloy fuel). The IFR fuels data were collected from irradiation of U-Zr based fuel alloy, with and without Pu additions, and clad in different types of steels, including HT9, D9, and 316 stainless-steel. Different types of data were generated during the program, and were based on the requirements associated with the DOE Advanced Liquid Metal Cooled Reactor (ALMR) program.

  12. A study of software safety analysis system for safety-critical software

    International Nuclear Information System (INIS)

    Chang, H. S.; Shin, H. K.; Chang, Y. W.; Jung, J. C.; Kim, J. H.; Han, H. H.; Son, H. S.

    2004-01-01

    The core factors and requirements for the safety-critical software traced and the methodology adopted in each stage of software life cycle are presented. In concept phase, Failure Modes and Effects Analysis (FMEA) for the system has been performed. The feasibility evaluation of selected safety parameter was performed and Preliminary Hazards Analysis list was prepared using HAZOP(Hazard and Operability) technique. And the check list for management control has been produced via walk-through technique. Based on the evaluation of the check list, activities to be performed in requirement phase have been determined. In the design phase, hazard analysis has been performed to check the safety capability of the system with regard to safety software algorithm using Fault Tree Analysis (FTA). In the test phase, the test items based on FMEA have been checked for fitness guided by an accident scenario. The pressurizer low pressure trip algorithm has been selected to apply FTA method to software safety analysis as a sample. By applying CASE tool, the requirements traceability of safety critical system has been enhanced during all of software life cycle phases

  13. Effect of a Background Noise on the Acoustic Leak Detection Methodology for a SFR Steam Generator

    International Nuclear Information System (INIS)

    Kim, Tae-Joon; Jeong, Ji-Young; Kim, Jong-Man

    2007-01-01

    The protection of a water/steam leak into a sodium in the SFR SG at an early phase of a leak origin depends on a fast response and sensitivity of a leak detection system not to a response against the several kinds of noises. The subject in this study is to introduce a detection performance by using our developed acoustic leak detection methodology discriminated by a backpropagation neural network according to a preprocessing of the 1/6 Octave band analysis or 1/12 Octave band analysis and the x n method defined by us. It was used for the acoustic signals generated from the simulation works which are the noises of an artificial background such as a scratching, a hammering on a steel structure and so on. In a previous study, we showed that the performance of a LabVIEW tool embedded with the developed acoustic leak detection methodology detected the SWR leak signals

  14. Safety analysis of autonomous excavator functionality

    International Nuclear Information System (INIS)

    Seward, D.; Pace, C.; Morrey, R.; Sommerville, I.

    2000-01-01

    This paper presents an account of carrying out a hazard analysis to define the safety requirements for an autonomous robotic excavator. The work is also relevant to the growing generic class of heavy automated mobile machinery. An overview of the excavator design is provided and the concept of a safety manager is introduced. The safety manager is an autonomous module responsible for all aspects of system operational safety, and is central to the control system's architecture. Each stage of the hazard analysis is described, i.e. system model creation, hazard definition and hazard analysis. Analysis at an early stage of the design process, and on a system that interfaces directly to an unstructured environment, exposes certain issues relevant to the application of current hazard analysis methods. The approach taken in the analysis is described. Finally, it is explained how the results of the hazard analysis have influenced system design, in particular, safety manager specifications. Conclusions are then drawn about the applicability of hazard analysis of requirements in general, and suggestions are made as to how the approach can be taken further

  15. Solid waste burial grounds interim safety analysis

    International Nuclear Information System (INIS)

    Saito, G.H.

    1994-01-01

    This Interim Safety Analysis document supports the authorization basis for the interim operation and restrictions on interim operations for the near-surface land disposal of solid waste in the Solid Waste Burial Grounds. The Solid Waste Burial Grounds Interim Safety Basis supports the upgrade progress for the safety analysis report and the technical safety requirements for the operations in the Solid Waste Burial Grounds. Accident safety analysis scenarios have been analyzed based on the significant events identified in the preliminary hazards analysis. The interim safety analysis provides an evaluation of the operations in the Solid Waste Burial Grounds to determine if the radiological and hazardous material exposures will be acceptable from an overall health and safety standpoint to the worker, the onsite personnel, the public, and the environment

  16. Solid waste burial grounds interim safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Saito, G.H.

    1994-10-01

    This Interim Safety Analysis document supports the authorization basis for the interim operation and restrictions on interim operations for the near-surface land disposal of solid waste in the Solid Waste Burial Grounds. The Solid Waste Burial Grounds Interim Safety Basis supports the upgrade progress for the safety analysis report and the technical safety requirements for the operations in the Solid Waste Burial Grounds. Accident safety analysis scenarios have been analyzed based on the significant events identified in the preliminary hazards analysis. The interim safety analysis provides an evaluation of the operations in the Solid Waste Burial Grounds to determine if the radiological and hazardous material exposures will be acceptable from an overall health and safety standpoint to the worker, the onsite personnel, the public, and the environment.

  17. Preliminary Acceptance Criteria for Safety Analysis of KALIMER-600 SFR

    International Nuclear Information System (INIS)

    Kwon, Young Min; Lee, Kwi Lim; Ha, Kwi Seok; Chang, Won Pyo; Jeong, Hae Yong

    2010-01-01

    The KALIMER-600 event categorization in the function of occurrence frequency has been made by traditional engineering judgment with information from some reference plants such as CRBR, PRISM and EFR. The dividing line between DBE and BDBE is the frequency of 10 -7 per plant-year. Each event belongs to one of five categories based upon its nominal frequency per reactor-year (f) as a criterion. (1) Moderate frequency Event (MF): f ≥ 10 -1 (2) Infrequent Event (IE): 10 -1 > f ≥ 10 -2 (3) Unlikely Event (UE): 10 -2 > f ≥ 10 -4 (4) Extremely Unlikely Event (XU): 10 -4 > f ≥ 10 -7 (5) Beyond DBE (BDBE): > f ≥ 10 -4

  18. Analysis of Phenix End-of-Life asymmetry test with multi-dimensional pool modeling of MARS-LMR code

    International Nuclear Information System (INIS)

    Jeong, H.-Y.; Ha, K.-S.; Choi, C.-W.; Park, M.-G.

    2015-01-01

    Highlights: • Pool behaviors under asymmetrical condition in an SFR were evaluated with MARS-LMR. • The Phenix asymmetry test was analyzed one-dimensionally and multi-dimensionally. • One-dimensional modeling has limitation to predict the cold pool temperature. • Multi-dimensional modeling shows improved prediction of stratification and mixing. - Abstract: The understanding of complicated pool behaviors and its modeling is essential for the design and safety analysis of a pool-type Sodium-cooled Fast Reactor. One of the remarkable recent efforts on the study of pool thermal–hydraulic behaviors is the asymmetrical test performed as a part of Phenix End-of-Life tests by the CEA. To evaluate the performance of MARS-LMR code, which is a key system analysis tool for the design of an SFR in Korea, in the prediction of thermal hydraulic behaviors during an asymmetrical condition, the Phenix asymmetry test is analyzed with MARS-LMR in the present study. Pool regions are modeled with two different approaches, one-dimensional modeling and multi-dimensional one, and the prediction results are analyzed to identify the appropriateness of each modeling method. The prediction with one-dimensional pool modeling shows a large deviation from the measured data at the early stage of the test, which suggests limitations to describe the complicated thermal–hydraulic phenomena. When the pool regions are modeled multi-dimensionally, the prediction gives improved results quite a bit. This improvement is explained by the enhanced modeling of pool mixing with the multi-dimensional modeling. On the basis of the results from the present study, it is concluded that an accurate modeling of pool thermal–hydraulics is a prerequisite for the evaluation of design performance and safety margin quantification in the future SFR developments

  19. Safety analysis procedures for PHWR

    International Nuclear Information System (INIS)

    Min, Byung Joo; Kim, Hyoung Tae; Yoo, Kun Joong

    2004-03-01

    The methodology of safety analyses for CANDU reactors in Canada, a vendor country, uses a combination of best-estimate physical models and conservative input parameters so as to minimize the uncertainty of the plant behavior predictions. As using the conservative input parameters, the results of the safety analyses are assured the regulatory requirements such as the public dose, the integrity of fuel and fuel channel, the integrity of containment and reactor structures, etc. However, there is not the comprehensive and systematic procedures for safety analyses for CANDU reactors in Korea. In this regard, the development of the safety analyses procedures for CANDU reactors is being conducted not only to establish the safety analyses system, but also to enhance the quality assurance of the safety assessment. In the first phase of this study, the general procedures of the deterministic safety analyses are developed. The general safety procedures are covered the specification of the initial event, selection of the methodology and accident sequences, computer codes, safety analysis procedures, verification of errors and uncertainties, etc. Finally, These general procedures of the safety analyses are applied to the Large Break Loss Of Coolant Accident (LBLOCA) in Final Safety Analysis Report (FSAR) for Wolsong units 2, 3, 4

  20. Homogeneous Minor Actinide Transmutation in SFR: Neutronic Uncertainties Propagation with Depletion

    International Nuclear Information System (INIS)

    Buiron, L.; Plisson-Rieunier, D.

    2015-01-01

    In the frame of next generation fast reactor design, the minimisation of nuclear waste production is one of the key objectives for current R and D. Among the possibilities studied at CEA, minor actinides multi-recycling is the most promising industrial way achievable in the near-term. Two main management options are considered: - Multi-recycling in a homogeneous way (minor actinides diluted in the driver fuel). If this solution can help achieving high transmutation rates, the negative impact of minor actinides on safety coefficients allows only a small fraction of the total heavy mass to be loaded in the core (∼ few %). - Multi-recycling in heterogeneous way by means of Minor Actinide Bearing Blanket (MABB) located at the core periphery. This solution offers more flexibility than the previous one, allowing a total minor actinides decoupled management from the core fuel. As the impact on feedback coefficient is small larger initial minor actinide mass can be loaded in this configuration. Starting from a breakeven Sodium Fast Reactor designed jointly by CEA, Areva and EdF teams, the so called SFR V2B, transmutation performances have been studied in frame on the French fleet for both options and various specific isotopic management (all minor actinides, americium only, etc.). Using these results, a sensitivity study has been performed to assess neutronic uncertainties (i.e coming from cross section) on mass balance on the most attractive configurations. This work in based on a new implementation of sensitivity on concentration with depletion in the ERANOS code package. Uncertainties on isotopes masses at the end of irradiation using various variance-covariance is discussed. (authors)

  1. Mapping of sodium void worth and doppler effect for sodium-cooled fast reactor - 15458

    International Nuclear Information System (INIS)

    Krepel, J.; Pelloni, S.; Bortot, S.; Panadero, A.L.; Mikityuk, K.

    2015-01-01

    The sodium-cooled fast reactor (SFR) represents the reference and the most technologically mastered system among the Generation-IV reactors. Nevertheless, the sodium void worth in the fuel regions of SFR is usually positive. To overcome this safety drawback, low-void sodium-cooled fast spectrum core (CFV) was proposed by CEA. Such a CFV core is used in the frame of WP6 'Core safety' of the FP7 Euratom ESNII+ project as a reference SFR design. The overall sodium void effect is negative for the CFV core. Nevertheless, locally it is positive in the fuel region and negative in the sodium plenum. Similarly, also the Doppler effect is spatially dependent and it varies between the inner and outer fuel regions and between the middle and lower blankets. Accordingly, knowledge of the local distributions or actually mappings of the two safety-related parameters will be necessary, before safety assessment and transient analysis can be done. In this study these maps have been produced using the deterministic code ERANOS. The obtained mapping shows strong local dependency of both safety-related effects. A sensitivity of the void effect to the sodium plenum modeling was also demonstrated. The results may serve as an input for the transient analysis of the CFV core or as a cross-check for the Monte Carlo method based maps. (authors)

  2. Methodology for a thermal analysis of a proposed SFR transport cask with the thermal code SYRTHES

    International Nuclear Information System (INIS)

    Peniguel, C.; Rupp, I.; Schneider, J. P.

    2010-01-01

    Fast reactors with liquid metal coolant have received a renewed interest owing to the need of a more efficient usage of the primary uranium resources, and they are one of the proposal for the next Generation IV. In the framework of the 2006 French law on sustainable management of radioactive materials and waste, an evaluation of the industrial perspectives of minor actinides transmutation advantages and drawbacks in Generation IV fast spectrum reactors system is requested for 2012. The CEA is in charge of studying the global problem, but on some aspects, EDF is interested to do its own exploratory studies. Among other points, transport is seen as important for the nuclear industry, to link points of production and treatment. Nuclear fuel is generally transported in thick walled rail or truck casks. These packages are designed to provide confinement, shielding and criticality protection during normal and severe transport conditions. Heat generated within the fuel (and a contribution of solar heating) makes the package becoming quite hot, but one must demonstrate that the cladding temperature does not exceed a long term temperature limit during normal transport. This paper presents a thermal study done on a package in which 9 SFR assemblies are included. Each of them is of hexagonal shape and contains 271 fuel pins. The approach followed for these calculations is to rely on an explicit representation of all pins. For these calculations a 2D analysis is performed thanks to the thermal code SYRTHES. Conduction is solved thanks to a finite element method, while thermal radiation is handled through a radiosity approach. The main aim of this paper is to present a possible numerical methodology to handle the thermal problem. (authors)

  3. SEISMIC ANALYSIS FOR PRECLOSURE SAFETY

    Energy Technology Data Exchange (ETDEWEB)

    E.N. Lindner

    2004-12-03

    The purpose of this seismic preclosure safety analysis is to identify the potential seismically-initiated event sequences associated with preclosure operations of the repository at Yucca Mountain and assign appropriate design bases to provide assurance of achieving the performance objectives specified in the Code of Federal Regulations (CFR) 10 CFR Part 63 for radiological consequences. This seismic preclosure safety analysis is performed in support of the License Application for the Yucca Mountain Project. In more detail, this analysis identifies the systems, structures, and components (SSCs) that are subject to seismic design bases. This analysis assigns one of two design basis ground motion (DBGM) levels, DBGM-1 or DBGM-2, to SSCs important to safety (ITS) that are credited in the prevention or mitigation of seismically-initiated event sequences. An application of seismic margins approach is also demonstrated for SSCs assigned to DBGM-2 by showing a high confidence of a low probability of failure at a higher ground acceleration value, termed a beyond-design basis ground motion (BDBGM) level. The objective of this analysis is to meet the performance requirements of 10 CFR 63.111(a) and 10 CFR 63.111(b) for offsite and worker doses. The results of this calculation are used as inputs to the following: (1) A classification analysis of SSCs ITS by identifying potential seismically-initiated failures (loss of safety function) that could lead to undesired consequences; (2) An assignment of either DBGM-1 or DBGM-2 to each SSC ITS credited in the prevention or mitigation of a seismically-initiated event sequence; and (3) A nuclear safety design basis report that will state the seismic design requirements that are credited in this analysis. The present analysis reflects the design information available as of October 2004 and is considered preliminary. The evolving design of the repository will be re-evaluated periodically to ensure that seismic hazards are properly

  4. SEISMIC ANALYSIS FOR PRECLOSURE SAFETY

    International Nuclear Information System (INIS)

    E.N. Lindner

    2004-01-01

    The purpose of this seismic preclosure safety analysis is to identify the potential seismically-initiated event sequences associated with preclosure operations of the repository at Yucca Mountain and assign appropriate design bases to provide assurance of achieving the performance objectives specified in the Code of Federal Regulations (CFR) 10 CFR Part 63 for radiological consequences. This seismic preclosure safety analysis is performed in support of the License Application for the Yucca Mountain Project. In more detail, this analysis identifies the systems, structures, and components (SSCs) that are subject to seismic design bases. This analysis assigns one of two design basis ground motion (DBGM) levels, DBGM-1 or DBGM-2, to SSCs important to safety (ITS) that are credited in the prevention or mitigation of seismically-initiated event sequences. An application of seismic margins approach is also demonstrated for SSCs assigned to DBGM-2 by showing a high confidence of a low probability of failure at a higher ground acceleration value, termed a beyond-design basis ground motion (BDBGM) level. The objective of this analysis is to meet the performance requirements of 10 CFR 63.111(a) and 10 CFR 63.111(b) for offsite and worker doses. The results of this calculation are used as inputs to the following: (1) A classification analysis of SSCs ITS by identifying potential seismically-initiated failures (loss of safety function) that could lead to undesired consequences; (2) An assignment of either DBGM-1 or DBGM-2 to each SSC ITS credited in the prevention or mitigation of a seismically-initiated event sequence; and (3) A nuclear safety design basis report that will state the seismic design requirements that are credited in this analysis. The present analysis reflects the design information available as of October 2004 and is considered preliminary. The evolving design of the repository will be re-evaluated periodically to ensure that seismic hazards are properly

  5. Safety balance: Analysis of safety systems; Bilans de surete: analyse par les organismes de surete

    Energy Technology Data Exchange (ETDEWEB)

    Delage, M; Giroux, C

    1990-12-01

    Safety analysis, and particularly analysis of exploitation of NPPs is constantly affected by EDF and by the safety authorities and their methodologies. Periodic safety reports ensure that important issues are not missed on daily basis, that incidents are identified and that relevant actions are undertaken. French safety analysis method consists of three principal steps. First type of safety balance is analyzed at the normal start-up phase for each unit including the final safety report. This enables analysis of behaviour of units ten years after their licensing. Second type is periodic operational safety analysis performed during a few years. Finally, the third step consists of safety analysis of the oldest units with the aim to improve the safety standards. The three steps of safety analysis are described in this presentation in detail with the aim to present the objectives and principles. Examples of most recent exercises are included in order to illustrate the importance of such analyses.

  6. Overview of Generation IV (Gen IV) Reactor Designs - Safety and Radiological Protection Considerations. Published on September 24, 2012

    International Nuclear Information System (INIS)

    Couturier, Jean; Bruna, Giovanni; Baudrand, Olivier; Blanc, Daniel; Ivanov, Evgeny; Bonneville, Herve; Clement, Bernard; Kissane, Martin; Meignen, Renaud; Monhardt, Daniel; Nicaise, Gregory; Bourgois, Thierry; Hache, Georges

    2012-01-01

    The purpose of this document is to provide an updated overview of specific safety and radiological protection issues for all the reactor concepts adopted by the GIF (Generation IV International Forum), independent of their advantages or disadvantages in terms of resource optimization or long-lived-waste reduction. In particular, this new document attempts to bring out the advantages and disadvantages of each concept in terms of safety, taking into account the Western European Nuclear Regulators' Association (WENRA) statement concerning safety objectives for new nuclear power plants. Using an identical framework for each reactor concept (sodium-cooled fast reactors or SFR, high / very-high temperature helium-cooled reactors of V/HTR, gas-cooled fast reactors or GFR, lead-or lead / bismuth-cooled fast reactors or LFR, molten salt reactors or MSR, and supercritical-water-cooled reactors or SCWR), this summary report provides some general conclusions regarding their safety and radiological protection issues, inspired by WENRA's safety objectives and on the basis of available information. Initial lessons drawn from the events at the Fukushima-Daiichi nuclear power plant in March 2011 have also been taken into account in IRSN's analysis of each reactor concept

  7. Review of the Conceptual Design for In-Vessel Fuel Handling Machines in SFR

    International Nuclear Information System (INIS)

    Kim, S. H.; Koo, G. H.

    2012-01-01

    The main in-vessel fuel handling machines in sodium cooled fast reactor(SFR) are composed of the in-vessel transfer machine(IVTM) and the rotating plug. These machines perform the function to handle fuel assemblies inside the reactor core during the refueling time. The IVTM should be able to access all areas above the reactor core and the fuel transfer port which can discharge the fuel assembly by the rotation of the rotating plug. In the 600 MWe demonstration reactor, the conceptual design of the in-vessel fuel handling machines was carried out. As shown in Fig. 1, the invessel fuel handling machines of the demonstration reactor are the double rotating plug type. With reference to the given core configuration of the demonstration reactor, the arrangement design of the rotating plug was carried out by using the developed simulation program. At present, the conceptual design of SFR prototype reactor which has small capacity of about 100 MWe is being started. Thus, it is necessary the economical efficiency and the reliability of the in-vessel fuel handling machines are reviewed according to the reduction of the power capacity. In this study, the preliminary design concepts of the main invessel fuel handling machines according to the fuel handling type are compared. Also, the design characteristics for the driving mechanism of the IVTM in the demonstration reactor and the recovery concept from the malfunction are reviewed

  8. Unprotected Accident Analyses of the 1200MWe GEN-IV Sodium-Cooled Fast Reactor Using the SSC-K Code

    International Nuclear Information System (INIS)

    Kwon, Young Min; Lee, Kwi Lim; Ha, Kwi Seok; Jeong, Hae Yong; Chang, Won Pyo; Seok, Su Dong; Lee, Yong Bum

    2010-02-01

    A conceptual design of an advanced breakeven sodium-cooled fast reactor (G4SFR) has recently been developed by KAERI under the national nuclear R and D plan. The G4SFR is a 1,200MWe metal-fueled pool-type sodium-cooled fast reactor adopting advanced safety design features. The G4SFR development plan focuses on particular technology development efforts to effectively meet the goals of the Generation-IV (GEN-IV) nuclear system such as efficient utilization of resources, economic competitiveness, a high standard of safety, and enhanced proliferation resistance. To enhance the safety of G4SFR, advanced design features of metal-fueled core, simple and large sodium-inventory primary heat transport system, and passive safety decay heat removal system are included in the reactor design. To evaluate potential safety characteristics of such advanced design features, the plant responses and safety margins were investigated using the system transient code SSC-K for three unprotected accidents of UTOP, ULOF, and ULOHS. It was shown that the G4SFR design has inherent and passive safety characteristics and is accommodating the selected ATWS events. The inherent safety mechanism of the reactor design makes the core shutdown with sufficient margin and passive removal of decay heat with matching the core power to heat sink by passive self-regulation. The self-regulation of power without scram is mainly due to the inherent negative reactivity feedback in conjunction with the large thermal inertia of the primary heat transport system and the passive decay heat removal. Such favorable inherent and passive safety behaviors of G4SFR are expected to virtually exclude the probability of severe accidents with potential for core damage

  9. Modelling the long-time stability of the engineered barriers of SFR with respect to climate changes

    International Nuclear Information System (INIS)

    Cronstrand, Peter

    2007-02-01

    Extensive modelling has been performed on various degradation processes that are related to climate changes and are believed to critically affect the long-term safety of the Silo and BMA at SFR. The model employs a variation of temperature and composition of infiltrating water, a dynamic update of the transport conditions parameterized through the porosity, fracture mechanics and addresses the expected freeze-thaw effects through a successive increase of the diffusivity. In order to evaluate the sensitivity of the model each effect is evaluated as a deviation from a base case before being incorporated in a complete model. Irrespective of the inclusion or exclusion of temperature and/or salinity variation and a dynamically updated diffusivity the calculations are in mutual agreement concerning the depth of the degradation, which for the Silo and BMA reaches half way through the concrete wall at year 100,000. A more substantial acceleration of the aging processes can be achieved by the inclusion of fractures or freeze-thaw effects, for which the results beyond year 50,000 are highly uncertain, but indicates that further degradation will be rapid. Although the present fracture model is incomplete, it demonstrates that rapid changes in transport conditions associated with emerging fractures can pose a serious threat to the repository

  10. Modelling the long-time stability of the engineered barriers of SFR with respect to climate changes

    Energy Technology Data Exchange (ETDEWEB)

    Cronstrand, Peter (Vattenfall Power Consultant, Stockholm (SE))

    2007-02-15

    Extensive modelling has been performed on various degradation processes that are related to climate changes and are believed to critically affect the long-term safety of the Silo and BMA at SFR. The model employs a variation of temperature and composition of infiltrating water, a dynamic update of the transport conditions parameterized through the porosity, fracture mechanics and addresses the expected freeze-thaw effects through a successive increase of the diffusivity. In order to evaluate the sensitivity of the model each effect is evaluated as a deviation from a base case before being incorporated in a complete model. Irrespective of the inclusion or exclusion of temperature and/or salinity variation and a dynamically updated diffusivity the calculations are in mutual agreement concerning the depth of the degradation, which for the Silo and BMA reaches half way through the concrete wall at year 100,000. A more substantial acceleration of the aging processes can be achieved by the inclusion of fractures or freeze-thaw effects, for which the results beyond year 50,000 are highly uncertain, but indicates that further degradation will be rapid. Although the present fracture model is incomplete, it demonstrates that rapid changes in transport conditions associated with emerging fractures can pose a serious threat to the repository

  11. SIMMER as a safety analysis tool

    International Nuclear Information System (INIS)

    Smith, L.L.; Bell, C.R.; Bohl, W.R.; Bott, T.F.; Dearing, J.F.; Luck, L.B.

    1982-01-01

    SIMMER has been used for numerous applications in fast reactor safety, encompassing both accident and experiment analysis. Recent analyses of transition-phase behavior in potential core disruptive accidents have integrated SIMMER testing with the accident analysis. Results of both the accident analysis and the verification effort are presented as a comprehensive safety analysis program

  12. Safety of ESWL in elderly: evaluation of independent predictors and comorbidity on stone-free rate and complications.

    Science.gov (United States)

    Polat, Fazli; Yeşil, Süleyman; Ak, Esat; Farahvash, Amirali; Karaoğlan, Ustünol; Biri, Hasan; Bozkirli, Ibrahim

    2012-07-01

    Urinary stone disease affects people of all ages. With its satisfactory efficacy ranges in all age groups and lack of side-effects, extracorporeal shock wave lithotripsy (ESWL) has become the preferred treatment modality for uncomplicated renal and proximal calculi ≤ 20 mm. In the present study, we aimed to assess the safety and efficacy of the ESWL treatment in elderly patients. A retrospective study was carried out on patients aged over 65 years who underwent shock wave lithotripsy at our Department from 2009 to 2011, with a Siemens Lithostar electromagnetic shockwave lithotripter. A total of 231 patients (157 males, 74 females) out of 1694 (13.6%) were studied. The patients were divided into two groups (group 1 = 65-70; group 2 >70). The effect of age and other possible predicting factors (sex, stone localization and stone size) were investigated. Concomitant diseases and related complications were also evaluated. An overall stone-free rate (SFR) of 82.2% was found. The influence of sex on SFR was non-significant. There was no significant difference when comparing SFR between the age groups. When patients were divided into those with renal and ureteral stones, the SFR were 94.4% and 67.6% (P 10 mm were 80% and 84.4%, respectively. Comorbidity was present in 148 patients. Complications were noted in 56 of 231 patients. Of 56 patients, 43 had minor complications and 13 major complications. ESWL seems to be an effective first-line treatment choice for urinary stones in elderly patients with careful patient selection and personalized preparation. © 2011 Japan Geriatrics Society.

  13. Development of safety analysis technology for integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sim, Suk K.; Song, J. H.; Chung, Y. J. and others

    1999-03-01

    Inherent safety features and safety system characteristics of the SMART integral reactor are investigated in this study. Performance and safety of the SMART conceptual design have been evaluated and confirmed through the performance and safety analyses using safety analysis system codes as well as a preliminary performance and safety analysis methodology. SMART design base events and their acceptance criteria are identified to develop a preliminary PIRT for the SMART integral reactor. Using the preliminary PIRT, a set of experimental program for the thermal hydraulic separate effect tests and the integral effect tests was developed for the thermal hydraulic model development and the system code validation. Safety characteristics as well as the safety issues of the integral reactor has been identified during the study, which will be used to resolve the safety issues and guide the regulatory criteria for the integral reactor. The results of the performance and safety analyses performed during the study were used to feedback for the SMART conceptual design. The performance and safety analysis code systems as well as the preliminary safety analysis methodology developed in this study will be validated as the SMART design evolves. The performance and safety analysis technology developed during the study will be utilized for the SMART basic design development. (author)

  14. Qualification of Simulation Software for Safety Assessment of Sodium Cooled Fast Reactors. Requirements and Recommendations

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Pointer, William David [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Sieger, Matt [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Flanagan, George F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Moe, Wayne [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); HolbrookINL, Mark [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-04-01

    The goal of this review is to enable application of codes or software packages for safety assessment of advanced sodium-cooled fast reactor (SFR) designs. To address near-term programmatic needs, the authors have focused on two objectives. First, the authors have focused on identification of requirements for software QA that must be satisfied to enable the application of software to future safety analyses. Second, the authors have collected best practices applied by other code development teams to minimize cost and time of initial code qualification activities and to recommend a path to the stated goal.

  15. A CRITICAL LOOK AT THE MASS-METALLICITY-STAR FORMATION RATE RELATION IN THE LOCAL UNIVERSE. I. AN IMPROVED ANALYSIS FRAMEWORK AND CONFOUNDING SYSTEMATICS

    Energy Technology Data Exchange (ETDEWEB)

    Salim, Samir; Salzer, John J. [Department of Astronomy, Indiana University, Bloomington, IN 47404 (United States); Lee, Janice C. [Space Telescope Science Institute, Baltimore, MD 21218 (United States); Ly, Chun [NASA Goddard Space Flight Center, Greenbelt, MD 20771 (United States); Brinchmann, Jarle [Leiden Observatory, Leiden University, NL-2300 RA Leiden (Netherlands); Davé, Romeel [University of the Western Cape, Bellville, Cape Town 7535 (South Africa); Dickinson, Mark [National Optical Astronomy Observatory, Tucson, AZ 85719 (United States); Charlot, Stéphane, E-mail: salims@indiana.edu [Institut d' Astrophysique de Paris, CNRS, F-75014 Paris (France)

    2014-12-20

    It has been proposed that the (stellar) mass-(gas) metallicity relation of galaxies exhibits a secondary dependence on star formation rate (SFR), and that the resulting M {sub *}-Z-SFR relation may be redshift-invariant, i.e., ''fundamental''. However, conflicting results on the character of the SFR dependence, and whether it exists, have been reported. To gain insight into the origins of the conflicting results, we (1) devise a non-parametric, astrophysically motivated analysis framework based on the offset from the star-forming ({sup m}ain{sup )} sequence at a given M {sub *} (relative specific SFR); (2) apply this methodology and perform a comprehensive re-analysis of the local M {sub *}-Z-SFR relation, based on SDSS, GALEX, and WISE data; and (3) study the impact of sample selection and of using different metallicity and SFR indicators. We show that metallicity is anti-correlated with specific SFR regardless of the indicators used. We do not find that the relation is spurious due to correlations arising from biased metallicity measurements or fiber aperture effects. We emphasize that the dependence is weak/absent for massive galaxies (log M {sub *} > 10.5), and that the overall scatter in the M {sub *}-Z-SFR relation does not greatly decrease from the M {sub *}-Z relation. We find that the dependence is stronger for the highest SSFR galaxies above the star-forming sequence. This two-mode behavior can be described with a broken linear fit in 12+log(O/H) versus log (SFR/M {sub *}), at a given M {sub *}. Previous parameterizations used for comparative analysis with higher redshift samples that do not account for the more detailed behavior of the local M {sub *}-Z-SFR relation may incorrectly lead to the conclusion that those samples follow a different relationship.

  16. Computer aided safety analysis 1989

    International Nuclear Information System (INIS)

    1990-04-01

    The meeting was conducted in a workshop style, to encourage involvement of all participants during the discussions. Forty-five (45) experts from 19 countries, plus 22 experts from the GDR participated in the meeting. A list of participants can be found at the end of this volume. Forty-two (42) papers were presented and discussed during the meeting. Additionally an open discussion was held on the possible directions of the IAEA programme on Computer Aided Safety Analysis. A summary of the conclusions of these discussions is presented in the publication. The remainder of this proceedings volume comprises the transcript of selected technical papers (22) presented in the meeting. It is the intention of the IAEA that the publication of these proceedings will extend the benefits of the discussions held during the meeting to a larger audience throughout the world. The Technical Committee/Workshop on Computer Aided Safety Analysis was organized by the IAEA in cooperation with the National Board for Safety and Radiological Protection (SAAS) of the German Democratic Republic in Berlin. The purpose of the meeting was to provide an opportunity for discussions on experiences in the use of computer codes used for safety analysis of nuclear power plants. In particular it was intended to provide a forum for exchange of information among experts using computer codes for safety analysis under the Technical Cooperation Programme on Safety of WWER Type Reactors (RER/9/004) and other experts throughout the world. A separate abstract was prepared for each of the 22 selected papers. Refs, figs tabs and pictures

  17. Subseabed disposal safety analysis

    International Nuclear Information System (INIS)

    Koplick, C.M.; Kabele, T.J.

    1982-01-01

    This report summarizes the status of work performed by Analytic Sciences Corporation (TASC) in FY'81 on subseabed disposal safety analysis. Safety analysis for subseabed disposal is divided into two phases: pre-emplacement which includes all transportation, handling, and emplacement activities; and long-term (post-emplacement), which is concerned with the potential hazard after waste is safely emplaced. Details of TASC work in these two areas are provided in two technical reports. The work to date, while preliminary, supports the technical and environmental feasibility of subseabed disposal of HLW

  18. 14 CFR 33.75 - Safety analysis.

    Science.gov (United States)

    2010-01-01

    ... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Safety analysis. 33.75 Section 33.75... STANDARDS: AIRCRAFT ENGINES Design and Construction; Turbine Aircraft Engines § 33.75 Safety analysis. (a... consequences of all failures that can reasonably be expected to occur. This analysis will take into account, if...

  19. 14 CFR 35.15 - Safety analysis.

    Science.gov (United States)

    2010-01-01

    ... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Safety analysis. 35.15 Section 35.15... STANDARDS: PROPELLERS Design and Construction § 35.15 Safety analysis. (a)(1) The applicant must analyze the.... This analysis will take into account, if applicable: (i) The propeller system in a typical installation...

  20. Safety analysis of spent fuel packaging

    International Nuclear Information System (INIS)

    Akamatsu, Hiroshi; Taniuchi, Hiroaki; Tai, Hideto

    1987-01-01

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

  1. Statistical considerations on safety analysis

    International Nuclear Information System (INIS)

    Pal, L.; Makai, M.

    2004-01-01

    The authors have investigated the statistical methods applied to safety analysis of nuclear reactors and arrived at alarming conclusions: a series of calculations with the generally appreciated safety code ATHLET were carried out to ascertain the stability of the results against input uncertainties in a simple experimental situation. Scrutinizing those calculations, we came to the conclusion that the ATHLET results may exhibit chaotic behavior. A further conclusion is that the technological limits are incorrectly set when the output variables are correlated. Another formerly unnoticed conclusion of the previous ATHLET calculations that certain innocent looking parameters (like wall roughness factor, the number of bubbles per unit volume, the number of droplets per unit volume) can influence considerably such output parameters as water levels. The authors are concerned with the statistical foundation of present day safety analysis practices and can only hope that their own misjudgment will be dispelled. Until then, the authors suggest applying correct statistical methods in safety analysis even if it makes the analysis more expensive. It would be desirable to continue exploring the role of internal parameters (wall roughness factor, steam-water surface in thermal hydraulics codes, homogenization methods in neutronics codes) in system safety codes and to study their effects on the analysis. In the validation and verification process of a code one carries out a series of computations. The input data are not precisely determined because measured data have an error, calculated data are often obtained from a more or less accurate model. Some users of large codes are content with comparing the nominal output obtained from the nominal input, whereas all the possible inputs should be taken into account when judging safety. At the same time, any statement concerning safety must be aleatory, and its merit can be judged only when the probability is known with which the

  2. Experimental and Numerical Analysis of S-CO2 Critical Flow for SFR Recovery System Design

    International Nuclear Information System (INIS)

    Kim, Min Seok; Jung, Hwa-Young; Ahn, Yoonhan; Lee, Jekyoung; Lee, Jeong Ik

    2016-01-01

    This paper presents both numerical and experimental studies of the critical flow of S-CO 2 while special attention is given to the turbo-machinery seal design. A computational critical flow model is described first. The experiments were conducted to validate the critical flow model. Various conditions have been tested to study the flow characteristic and provide validation data for the model. The comparison of numerical and experimental results of S-CO 2 critical flow will be presented. In order to eliminate SWR, a concept of coupling the Supercritical CO 2 (S-CO 2 ) cycle with SFR has been proposed. It is known that for a closed system controlling the inventory is important for stable operation and achieving high efficiency. Since the S-CO 2 power cycle is a highly pressurized system, certain amount of leakage flow is inevitable in the rotating turbo-machinery via seals. To simulate the CO 2 leak flow in a turbo-machinery with higher accuracy in the future, the real gas effect and friction factor will be considered for the CO 2 critical flow model. Moreover, experimentally obtained temperature data were somewhat different from the numerically obtained temperature due to the insufficient insulation and large thermal inertia of the CO 2 critical flow facility. Insulation in connecting pipes and the low-pressure tank will be added and additional tests will be conducted

  3. Conceptual Design on Primary Control Rod Drive Mechanism of a Prototype Gen-IV SFR

    International Nuclear Information System (INIS)

    Lee, Jae Han; Koo, Gyeong Hoi

    2013-01-01

    This paper describes the key concept of the drive mechanism, and suggests a required motor power and reducer gears to meet the functional design requirements, and a seismic response analysis of CRDM housing is performed to check its structural integrity. An AC servo motor is selected as a CRA driving power because it uses permanent magnets and is brushless type while DC motor needs a brush and a coil rotates. The control shim motor size is constrained by a housing diameter of 250mm. The driving system has several design requirements. To calculate the motor power, the drive shaft torque is needed. One part of the drive shaft has a lead screw, driving by a ball-nut. The ball screw driver torque (Tr) is calculated by some equations as follow; A servo motor with a nominal power of 100W, a nominal torque of 0.32 N-m (max. 0.48N-m) is selected considering a safety margin. Its diameter is about 50mm. The fast drive-in motor needs a strong power to insert enforcedly the stuck CRA into core within a required time. The motor sizes are calculated by the same procedure. The diameters are in the range of 80mm to 110mm by the insertion time (10 ∼ 24 seconds). The prototype Gen-IV SFR (sodium-cooled Fast Reactor) is of 150MWe capacity. The reactor has six primary control rod assemblies(CRAs). The primary control rod is used for power control, burn-up compensation and reactor shutdown in response to demands from the plant control or protection systems. The control rod drive mechanism (CRDM) consists of the drive motor assembly, the driveline, and its housing. The driveline consists of three concentric members of a drive shaft, a tension tube, and a position indicator rod, and it connects the drive motor assembly to the CRA. Main issue is that these many driving parts shall be enclosed within a limited housing diameter because the available pitch of CRDMs is limited by 300mm

  4. Conceptual Design on Primary Control Rod Drive Mechanism of a Prototype Gen-IV SFR

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Han; Koo, Gyeong Hoi [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    This paper describes the key concept of the drive mechanism, and suggests a required motor power and reducer gears to meet the functional design requirements, and a seismic response analysis of CRDM housing is performed to check its structural integrity. An AC servo motor is selected as a CRA driving power because it uses permanent magnets and is brushless type while DC motor needs a brush and a coil rotates. The control shim motor size is constrained by a housing diameter of 250mm. The driving system has several design requirements. To calculate the motor power, the drive shaft torque is needed. One part of the drive shaft has a lead screw, driving by a ball-nut. The ball screw driver torque (Tr) is calculated by some equations as follow; A servo motor with a nominal power of 100W, a nominal torque of 0.32 N-m (max. 0.48N-m) is selected considering a safety margin. Its diameter is about 50mm. The fast drive-in motor needs a strong power to insert enforcedly the stuck CRA into core within a required time. The motor sizes are calculated by the same procedure. The diameters are in the range of 80mm to 110mm by the insertion time (10 ∼ 24 seconds). The prototype Gen-IV SFR (sodium-cooled Fast Reactor) is of 150MWe capacity. The reactor has six primary control rod assemblies(CRAs). The primary control rod is used for power control, burn-up compensation and reactor shutdown in response to demands from the plant control or protection systems. The control rod drive mechanism (CRDM) consists of the drive motor assembly, the driveline, and its housing. The driveline consists of three concentric members of a drive shaft, a tension tube, and a position indicator rod, and it connects the drive motor assembly to the CRA. Main issue is that these many driving parts shall be enclosed within a limited housing diameter because the available pitch of CRDMs is limited by 300mm.

  5. Software FMEA analysis for safety-related application software

    International Nuclear Information System (INIS)

    Park, Gee-Yong; Kim, Dong Hoon; Lee, Dong Young

    2014-01-01

    Highlights: • We develop a modified FMEA analysis suited for applying to software architecture. • A template for failure modes on a specific software language is established. • A detailed-level software FMEA analysis on nuclear safety software is presented. - Abstract: A method of a software safety analysis is described in this paper for safety-related application software. The target software system is a software code installed at an Automatic Test and Interface Processor (ATIP) in a digital reactor protection system (DRPS). For the ATIP software safety analysis, at first, an overall safety or hazard analysis is performed over the software architecture and modules, and then a detailed safety analysis based on the software FMEA (Failure Modes and Effect Analysis) method is applied to the ATIP program. For an efficient analysis, the software FMEA analysis is carried out based on the so-called failure-mode template extracted from the function blocks used in the function block diagram (FBD) for the ATIP software. The software safety analysis by the software FMEA analysis, being applied to the ATIP software code, which has been integrated and passed through a very rigorous system test procedure, is proven to be able to provide very valuable results (i.e., software defects) that could not be identified during various system tests

  6. Economic Analysis of Different Nuclear Fuel Cycle Options

    International Nuclear Information System (INIS)

    Ko, W.; Gao, F.

    2012-01-01

    An economic analysis has been performed to compare four nuclear fuel cycle options: a once-through cycle (OT), DUPIC recycling, thermal recycling using MOX fuel in a pressurized water reactor (PWR-MOX), and sodium fast reactor recycling employing pyro processing (Pyro-SFR). This comparison was made to suggest an economic competitive fuel cycle for the Republic of Korea. The fuel cycle cost (FCC) has been calculated based on the equilibrium material flows integrated with the unit cost of the fuel cycle components. The levelized fuel cycle costs (LFCC) have been derived in terms of mills/kWh for a fair comparison among the FCCs, and the results are as follows: OT 7.35 mills/kWh, DUPIC 9.06 mills/kWh, PUREX-MOX 8.94 mills/kWh, and Pyro-SFR 7.70 mills/kWh. Due to unavoidable uncertainties, a cost range has been applied to each unit cost, and an uncertainty study has been performed accordingly. A sensitivity analysis has also been carried out to obtain the break-even uranium price (215$/kgU) for the Pyro-SFR against the OT, which demonstrates that the deployment of the Pyro-SFR may be economical in the foreseeable future. The influence of pyro techniques on the LFCC has also been studied to determine at which level the potential advantages of Pyro-SFR can be realized.

  7. Gas-cooled reactor safety and accident analysis

    International Nuclear Information System (INIS)

    1985-12-01

    The Specialists' Meeting on Gas-Cooled Reactor Safety and Accident Analysis was convened by the International Atomic Energy Agency in Oak Ridge on the invitation of the Department of Energy in Washington, USA. The meeting was hosted by the Oak Ridge National Laboratory. The purpose of the meeting was to provide an opportunity to compare and discuss results of safety and accident analysis of gas-cooled reactors under development, construction or in operation, to review their lay-out, design, and their operational performance, and to identify areas in which additional research and development are needed. The meeting emphasized the high safety margins of gas-cooled reactors and gave particular attention to the inherent safety features of small reactor units. The meeting was subdivided into four technical sessions: Safety and Related Experience with Operating Gas-Cooled Reactors (4 papers); Risk and Safety Analysis (11 papers); Accident Analysis (9 papers); Miscellaneous Related Topics (5 papers). A separate abstract was prepared for each of these papers

  8. Safety analysis - current and future regulatory challenges

    Energy Technology Data Exchange (ETDEWEB)

    Jamieson, T., E-mail: Terry.Jamieson@cnsc-ccsn.gc.ca [Canadian Nuclear Safety Commission, Ottawa, ON (Canada)

    2015-07-01

    'Full text:' The current and future regulatory challenges associated with deterministic safety analysis are reviewed, including: 1. The CNSC's and safety control areas. 2. Traditional safety analysis approach. 3. Experience gained and impact. 4. Current analysis and regulatory approaches. 5. Current status. 6. Complexity and challenges In particular, the technical, regulatory and strategic aspects of these challenges are discussed. (author)

  9. Safety analysis - current and future regulatory challenges

    International Nuclear Information System (INIS)

    Jamieson, T.

    2015-01-01

    'Full text:' The current and future regulatory challenges associated with deterministic safety analysis are reviewed, including: 1. The CNSC's and safety control areas. 2. Traditional safety analysis approach. 3. Experience gained and impact. 4. Current analysis and regulatory approaches. 5. Current status. 6. Complexity and challenges In particular, the technical, regulatory and strategic aspects of these challenges are discussed. (author)

  10. Effects of Ta addition on the Microstructural and Mechanical Properties of 9Cr-0.5Mo-2W F/M Steels for a SFR Fuel Cladding

    International Nuclear Information System (INIS)

    Baek, Jong-Hyuk; Han, Chang-Hee; Kim, Tae-Kyu; Kim, Sung-Ho; Lee, Chan-Bock

    2007-01-01

    Today twenty fission reactors provide about 40% of the domestic electricity supply. The world-wide distribution of some nuclear reactors will be aging and will need replacement and enhancement to both keep pace with and to take up a large share of the growing world-wide electricity demand. A new generation (Gen IV) of nuclear plant concepts has become the focus of international advanced reactor activity. Gen IV nuclear systems embodies greater improvements and innovative advances in technology over earlier ones. The Gen IV systems are to have a considerable increase in safety and be economically competitive when compared with the existed commercial reactors. In particular, the systems should produce a significantly reduced volume of nuclear wastes. From this point of view, sodium-cooled Fast Reactor (SFR) is strongly considered as a future nuclear energy system in Korea

  11. Acoustic displacement sensor for harsh environment: application to SFR core support plate monitoring

    International Nuclear Information System (INIS)

    PeRISSE, J.; MACe, J.R.; VOUAGNER, P.

    2013-06-01

    The need for instrumentation able to monitor internal parameters inside reactor vessels during plant operation is getting stronger. Internal mechanical structures important for safety are concerned: for example core support plate, fuel assemblies or primary pumps. Because of very harsh environmental conditions (high temperature, pressure and radiation) and maintenance requirements, sensors are generally located on the outer shell of the vessel with, for example, strain gages, accelerometers, eddy current or US sensors. Then, some complex signal processing calculations must be performed to address internal structure behavior or health analysis but with bias effects (transfer path analysis method for example). This study will show an original displacement sensor based on an acoustic wave guide that can measure small displacement of mechanical structures inside reactor vessels. The application selected in this case is the monitoring of the core support plate for a sodium fast reactor (SFR). The wave guide - a thin tube sealed with pressurized argon gas inside - is installed inside the liquid sodium vessel (temperature between 400 deg. C to 550 deg. C). One extremity is connected to the mechanical structure for control. It includes two acoustic reflectors; such reflectors are dedicated to a calibration procedure to estimate the acoustic wave velocity whatever the temperature profile along the wave guide (velocity is temperature dependent). The opposite extremity of the wave guide is located outside the vessel and includes an emission/reception acoustic transducer. Using acoustic pulse reflectometry method, a plane wave pressure signal propagates inside the tube and reflects from the extremity and acoustic reflectors. The pulse-echo signals are recorded and processed in the frequency domain. Signal processing is performed to estimate the time of flight of pulse reflections patterns along the acoustic path. Then, monitored structure displacement - i.e. movement of the

  12. Development of safety analysis technology for LMR

    International Nuclear Information System (INIS)

    Hahn, Do Hee; Kwon, Y. M.; Kim, K. D.

    2000-05-01

    The analysis methodologies as well as the analysis computer code system for the transient, HCDA, and containment performance analyses, which are required for KALIMER safety analyses, have been developed. The SSC-K code has been developed based on SSC-L which is an analysis code for loop type LMR, by improving models necessary for the KALIMER system analysis, and additional models have been added to the code. In addition, HCDA analysis model has been developed and the containment performance analysis code has been also improved. The preliminary basis for the safety analysis has been established, and the preliminary safety analyses for the key design features have been performed. In addition, a state-of-art analysis for LMR PSA and overseas safety and licensing requirements have been reviewed. The design database for the systematic management of the design documents as well as design processes has been established as well

  13. Development of safety analysis technology for LMR

    Energy Technology Data Exchange (ETDEWEB)

    Hahn, Do Hee; Kwon, Y. M.; Kim, K. D. [and others

    2000-05-01

    The analysis methodologies as well as the analysis computer code system for the transient, HCDA, and containment performance analyses, which are required for KALIMER safety analyses, have been developed. The SSC-K code has been developed based on SSC-L which is an analysis code for loop type LMR, by improving models necessary for the KALIMER system analysis, and additional models have been added to the code. In addition, HCDA analysis model has been developed and the containment performance analysis code has been also improved. The preliminary basis for the safety analysis has been established, and the preliminary safety analyses for the key design features have been performed. In addition, a state-of-art analysis for LMR PSA and overseas safety and licensing requirements have been reviewed. The design database for the systematic management of the design documents as well as design processes has been established as well.

  14. Preliminary Assessment of PHTS Pump Piping Break Accident of DSFR-600

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Andong; Bae, Moohoon; Choi, Yongwon; Suh, Namduk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2013-05-15

    KINS is evaluating the applicability of TRACE code for safety analysis of SFR Since 2012. Based on the steady-state input deck for Demonstration Sodium Cooled Fast Reactor 600MW (DSFR-600) component-wise specific modeling is developed for DSFR-600. Preliminary analysis was performed with TRACE code for DSFR-600 PHTS pump piping break accident. The calculation result showed that the calculated safety parameters are conforms to the design criteria for DBA accidents. RHRS design of DSFR-600 and its performance during transient was also reviewed by sensitivity study on the effect of sodium condition to the transient decay heat removal capability of RHRS. Following insights are identified. These should be considered in improving the design also in licensing review of SFR safety analysis. The transient performance of RHRS might differ from the component's design capacity. RHRS's transient performance also should be included in the design documents and validated with reasonable test and/or analysis with consideration of the variation of coolant conditions during transient. The analytic model used for safety analysis should consider 3-D effect of vessel pool and its uncertainty with reasonable conservatism.

  15. Safety analysis reports - new strategies

    International Nuclear Information System (INIS)

    Booth, J.A.

    1994-01-01

    Within the past year there have been many external changes in the requirements of safety analysis reports. Now there is emphasis on open-quotes graded approachesclose quotes depending on the Hazard Classification of the project. The Energy Facility Contractors Group (EFCOG) has a Safety Analysis Working Group. The results of this group for the past year are discussed as well as the implications for EG ampersand G. New strategies include ideas for incorporating the graded approach, auditable safety documents, additional guidance for Hazard Classification per DOE-STD-1027-92. The emphasis in the paper is on those projects whose hazard classification is category three or less

  16. Deep Borehole Disposal Safety Analysis.

    Energy Technology Data Exchange (ETDEWEB)

    Freeze, Geoffrey A. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Stein, Emily [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Price, Laura L. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); MacKinnon, Robert J. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Tillman, Jack Bruce [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States)

    2016-10-01

    This report presents a preliminary safety analysis for the deep borehole disposal (DBD) concept, using a safety case framework. A safety case is an integrated collection of qualitative and quantitative arguments, evidence, and analyses that substantiate the safety, and the level of confidence in the safety, of a geologic repository. This safety case framework for DBD follows the outline of the elements of a safety case, and identifies the types of information that will be required to satisfy these elements. At this very preliminary phase of development, the DBD safety case focuses on the generic feasibility of the DBD concept. It is based on potential system designs, waste forms, engineering, and geologic conditions; however, no specific site or regulatory framework exists. It will progress to a site-specific safety case as the DBD concept advances into a site-specific phase, progressing through consent-based site selection and site investigation and characterization.

  17. Establishment of the international collaboration and licensing preparation planning for the specific design of a prototype SFR

    International Nuclear Information System (INIS)

    Kim, Y. G.; Joo, H. K.; Cho, C. H.; Yoo, J. W.; Lee, D. U.; Ahn, K. S.; Hwang, Y. S.

    2013-05-01

    The conceptual design of prototype of Gen IV SFR (PGSFR) will be early determined through the review of the international experts. After this, the technology demonstration plan and validation of fuel design will be determined in more detail. The project will be accomplished efficiently by introducing the proven technology already validated from the international collaboration. The conceptual design and its requirements of PGSFR will be reviewed by ANL, who has a lot of design experiences in the metal fueled SFR development. The collaboration with ANL has been done through Work For Others (WFO) contract, and the MOU was signed between SFRA and Terra Power(USA), and SFRA and IGCAR. The licensing issues raised during PFBR and FBTR licensing in India will be discussed and reflected into the PGSFR design by inviting the high level expert from India, for example Dr. Chetal in IGCAR. The specific design, technology validation plan and fuel development plan will be established in more detail through the annual International Technical Review Meeting (ITRM) and experimental facilities available from the international institute and companies, which will be the basis for shortening the project period and to reduce the development cost

  18. National Waste Repository Novi Han operational safety analysis report. Safety assessment methodology

    International Nuclear Information System (INIS)

    2003-01-01

    The scope of the safety assessment (SA), presented includes: waste management functions (acceptance, conditioning, storage, disposal), inventory (current and expected in the future), hazards (radiological and non-radiological) and normal and accidental modes. The stages in the development of the SA are: criteria selection, information collection, safety analysis and safety assessment documentation. After the review the facilities functions and the national and international requirements, the criteria for safety level assessment are set. As a result from the 2nd stage actual parameters of the facility, necessary for safety analysis are obtained.The methodology is selected on the base of the comparability of the results with the results of previous safety assessments and existing standards and requirements. The procedure and requirements for scenarios selection are described. A radiological hazard categorisation of the facilities is presented. Qualitative hazards and operability analysis is applied. The resulting list of events are subjected to procedure for prioritization by method of 'criticality analysis', so the estimation of the risk is given for each event. The events that fall into category of risk on the boundary of acceptability or are unacceptable are subjected to the next steps of the analysis. As a result the lists with scenarios for PSA and possible design scenarios are established. PSA logical modeling and quantitative calculations of accident sequences are presented

  19. NPP Temelin safety analysis reports and PSA status

    International Nuclear Information System (INIS)

    Mlady, O.

    1999-01-01

    To enhance the safety level of Temelin NPP, recommendations of the international reviews were implemented into the design as well as into organization of the plant construction and preparation for operation. The safety assessment of these design changes has been integrated and reflected in the Safety Analysis Reports, which follow the internationally accepted guidelines. All safety analyses within Safety Analysis Reports were repeated carefully considering technical improvements and replacements to complement preliminary safety documentation. These analyses were performed by advanced western computer codes to the depth and in the structure required by western standards. The Temelin NPP followed a systematic approach in the functional design of the Reactor Protection System and related safety analyses. Modifications of reactor protection system increase defense in depth and facilitate demonstrating that LOCA and radiological limits are met for non-LOCA events. The rigorous safety analysis methodology provides assurance that LOCA and radiological limits are met. Established and accepted safety analysis methodology and accepted criteria were applied to Temelin NPP meeting US NRC and Czech Republic requirements. IAEA guidelines and recommendations

  20. Identification of important phenomena under sodium fire accidents based on PIRT process with factor analysis in sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Aoyagi, Mitsuhiro; Uchibori, Akihiro; Kikuchi, Shin; Takata, Takashi; Ohno, Shuji; Ohshima, Hiroyuki

    2016-01-01

    The PIRT (Phenomena Identification and Ranking Table) process is an effective method to identify key phenomena involved in safety issues in nuclear power plants. The present PIRT process is aimed to validate sodium fire analysis codes. Because a sodium fire accident in sodium-cooled fast reactor (SFR) involves complex phenomena, various figures of merit (FOMs) could exist in this PIRT process. In addition, importance evaluation of phenomena for each FOM should be implemented in an objective manner under the PIRT process. This paper describes the methodology for specification of FOMs, identification of associated phenomena and importance evaluation of each associated phenomenon in order to complete a ranking table of important phenomena involved in a sodium fire accident in an SFR. The FOMs were specified through factor analysis in this PIRT process. Physical parameters to be quantified by a sodium fire analysis code were identified by considering concerns resulting from sodium fire in the factor analysis. Associated phenomena were identified through the element- and sequence-based phenomena analyses as is often conducted in PIRT processes. Importance of each associated phenomenon was evaluated by considering the sequence-based analysis of associated phenomena correlated with the FOMs. Then, we complete the ranking table through the factor and phenomenon analyses. (author)

  1. System safety engineering analysis handbook

    Science.gov (United States)

    Ijams, T. E.

    1972-01-01

    The basic requirements and guidelines for the preparation of System Safety Engineering Analysis are presented. The philosophy of System Safety and the various analytic methods available to the engineering profession are discussed. A text-book description of each of the methods is included.

  2. Current status of safety analysis report for ANPP

    International Nuclear Information System (INIS)

    Amirjanyan, A.

    1999-01-01

    Current situation concerning Armenian NPP safety analysis report is considered within the frame of accepted safety practice. Licensing procedure is being developed. Technical support group was established in the Armenian Nuclear Regulatory Authority (ANRA). The task of the group is to study modern methods of NPP in depth safety analysis for technical assistance for the ANRA, and perform independent safety assessments. ANRA will be obliged to demand assistance from various foreign organisations for preparation of different parts of the Safety Analysis Report like determination though certain parts can be prepared in Armenia

  3. Review of the Safety Design Approaches in Sodium Fast Reactors

    International Nuclear Information System (INIS)

    Suk, Soo Dong; Lee, Yong Bum

    2009-12-01

    The principle of the Defense in depth is essential in securing the safety of nuclear power plants, that is, to prevent cores-damaging severs accidents and to minimize the radiological consequences of the accidents 'as low as possible' (ALARA). One of the major design features of sodium fast reactors (SFRs) is that it has a large amount of sodium in the reactor vessel, providing a large heat capacity, such that it is feasible to contain the consequences of sever core damaging accidents in the vessel and primary system boundary. Containment of a severe accident in the primary system boundary, that is called in-vessel retention(IVR), is not a licensing requirement but set up as a design goal in most of the SFR design in the context of risk minimization. The objective of this report is to broadly review and compare the approaches and efforts made in the some of the major SFR designs of the US, Europe and Japan to prevent severe accidents and mitigate their consequences should they occur. Specifically, the subjects described in this report include design criteria or requirements, accident categorization and acceptance criteria, design features to prevent and contain severs accidents

  4. NKS/SOS-1 seminar on safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Lauridsen, K. [Risoe National Lab., Roskilde (Denmark); Anderson, K. [Karinta-Konsult (Sweden); Pulkkinen, U. [VTT Automation (Finland)

    2001-05-01

    The report describes presentations and discussions at a seminar held at Risoe on March 22-23, 2000. The title of the seminar was NKS/SOS-1 - Safety Analysis. It dealt with issues of relevance for the safety analysis for the entire nuclear safety field (notably reactors and nuclear waste repositories). Such issues were: objectives of safety analysis, risk criteria, decision analysis, expert judgement and risk communication. In addition, one talk dealt with criteria for chemical industries in Europe. The seminar clearly showed that the concept of risk is multidimensional, which makes clarity and transparency essential elements in risk communication, and that there are issues of common concern between different applications, such as how to deal with different kinds of uncertainty and expert judgement. (au)

  5. A neutronics study for improving the safety and performance parameters of a 3600 MWth Sodium-cooled Fast Reactor

    International Nuclear Information System (INIS)

    Sun, Kaichao; Krepel, Jiri; Mikityuk, Konstantin; Chawla, Rakesh

    2013-01-01

    Highlights: ► The potential for neutronics design optimization is assessed for a large SFR core. ► Both beginning-of-life and equilibrium fuel cycle conditions are considered. ► The sodium void effect is decomposed via a neutron balance based methodology. ► The optimized core options adopt an appropriate sodium plenum design to reduce the void effect. ► The introduction of moderator pins is considered for enhancing the Doppler effect. - Abstract: The Sodium-cooled Fast Reactor (SFR) is one of the most promising Generation IV systems with many performance advantages, but has one dominating neutronics drawback – a positive sodium void reactivity. The starting point for the present study is an SFR core design considered in the Collaborative Project on the European Sodium-cooled Fast Reactor (CP-ESFR). The aim is to analyze, for this reference core, four safety and performance parameters from the viewpoint of four different optimization options, and to propose possible optimized core designs. In doing so, the study focuses not only on the beginning-of-life state of the core, but also on the beginning of equilibrium closed fuel cycle. The four studied optimization options are: (a) introducing an upper sodium plenum and boron layer, (b) varying the core height-to-diameter (H/D) ratio, (c) introducing moderator pins into the fuel assembly, and (d) modifying the initial plutonium content. The sensitivity of the void reactivity, Doppler constant, nominal reactivity and breeding gain has been evaluated. In particular, the void reactivity, which is the most crucial safety parameter for the SFR, has been decomposed into its reaction-wise, isotope-wise and energy-group-wise components using a methodology based on the neutron balance equation. Extended voiding in the upper sodium plenum region – in conjunction with the effect of a boron layer introduced above the plenum – is found to be particularly effective in the void effect reduction while, at the same time

  6. HANFORD SAFETY ANALYSIS & RISK ASSESSMENT HANDBOOK (SARAH)

    Energy Technology Data Exchange (ETDEWEB)

    EVANS, C B

    2004-12-21

    The purpose of the Hanford Safety Analysis and Risk Assessment Handbook (SARAH) is to support the development of safety basis documentation for Hazard Category 2 and 3 (HC-2 and 3) U.S. Department of Energy (DOE) nuclear facilities to meet the requirements of 10 CFR 830, ''Nuclear Safety Management''. Subpart B, ''Safety Basis Requirements.'' Consistent with DOE-STD-3009-94, Change Notice 2, ''Preparation Guide for U.S. Department of Energy Nonreactor Nuclear Facility Documented Safety Analyses'' (STD-3009), and DOE-STD-3011-2002, ''Guidance for Preparation of Basis for Interim Operation (BIO) Documents'' (STD-3011), the Hanford SARAH describes methodology for performing a safety analysis leading to development of a Documented Safety Analysis (DSA) and derivation of Technical Safety Requirements (TSR), and provides the information necessary to ensure a consistently rigorous approach that meets DOE expectations. The DSA and TSR documents, together with the DOE-issued Safety Evaluation Report (SER), are the basic components of facility safety basis documentation. For HC-2 or 3 nuclear facilities in long-term surveillance and maintenance (S&M), for decommissioning activities, where source term has been eliminated to the point that only low-level, residual fixed contamination is present, or for environmental remediation activities outside of a facility structure, DOE-STD-1120-98, ''Integration of Environment, Safety, and Health into Facility Disposition Activities'' (STD-1120), may serve as the basis for the DSA. HC-2 and 3 environmental remediation sites also are subject to the hazard analysis methodologies of this standard.

  7. Safety analysis reports. Current status (third key report)

    International Nuclear Information System (INIS)

    1999-01-01

    A review of Ukrainian regulations and laws concerned with Nuclear power and radiation safety is presented with an overview of the requirements for the Safety Analysis Report Contents. Status of Safety Analysis Reports (SAR) is listed for each particular Ukrainian NPP including SAR development schedules. Organisational scheme of SAR development works includes: general technical co-ordination on Safety Analysis Report development; list of leading organisations and utilization of technical support within international projects

  8. Status of Ignalina's safety analysis reports

    International Nuclear Information System (INIS)

    Uspuras, E.

    1999-01-01

    Ignalina NPP is unique among RBMK type reactors in the scope and comprehensiveness of international studies which have been performed to verify its design parameters and analyze risk levels. International assistance took several forms, a very valuable mod of assistance utilized the knowledge of international experts in extensive international studies whose purpose was: collection, systematization and verification of plant design data; analysis of risk levels; recommendations leading to improvements in the safety lave; transfer of state of the art analytical methodology to Lithuanian specialists. The major large scale international studies include: probabilistic risk analysis; extensive international study meant to provide comprehensive overview of plant status with special emphasis on safety aspects; an extensive review of the Safety Analysis Report by an independent group of international experts. In spite of the safety improvements and analyses which have been performed at the Ignalina NPP, much remains to be done in the nearest future

  9. Hot Cell Facility (HCF) Safety Analysis Report

    Energy Technology Data Exchange (ETDEWEB)

    MITCHELL,GERRY W.; LONGLEY,SUSAN W.; PHILBIN,JEFFREY S.; MAHN,JEFFREY A.; BERRY,DONALD T.; SCHWERS,NORMAN F.; VANDERBEEK,THOMAS E.; NAEGELI,ROBERT E.

    2000-11-01

    This Safety Analysis Report (SAR) is prepared in compliance with the requirements of DOE Order 5480.23, Nuclear Safety Analysis Reports, and has been written to the format and content guide of DOE-STD-3009-94 Preparation Guide for U. S. Department of Energy Nonreactor Nuclear Safety Analysis Reports. The Hot Cell Facility is a Hazard Category 2 nonreactor nuclear facility, and is operated by Sandia National Laboratories for the Department of Energy. This SAR provides a description of the HCF and its operations, an assessment of the hazards and potential accidents which may occur in the facility. The potential consequences and likelihood of these accidents are analyzed and described. Using the process and criteria described in DOE-STD-3009-94, safety-related structures, systems and components are identified, and the important safety functions of each SSC are described. Additionally, information which describes the safety management programs at SNL are described in ancillary chapters of the SAR.

  10. Approach to uncertainty evaluation for safety analysis

    International Nuclear Information System (INIS)

    Ogura, Katsunori

    2005-01-01

    Nuclear power plant safety used to be verified and confirmed through accident simulations using computer codes generally because it is very difficult to perform integrated experiments or tests for the verification and validation of the plant safety due to radioactive consequence, cost, and scaling to the actual plant. Traditionally the plant safety had been secured owing to the sufficient safety margin through the conservative assumptions and models to be applied to those simulations. Meanwhile the best-estimate analysis based on the realistic assumptions and models in support of the accumulated insights could be performed recently, inducing the reduction of safety margin in the analysis results and the increase of necessity to evaluate the reliability or uncertainty of the analysis results. This paper introduces an approach to evaluate the uncertainty of accident simulation and its results. (Note: This research had been done not in the Japan Nuclear Energy Safety Organization but in the Tokyo Institute of Technology.) (author)

  11. Hot Cell Facility (HCF) Safety Analysis Report

    International Nuclear Information System (INIS)

    MITCHELL, GERRY W.; LONGLEY, SUSAN W.; PHILBIN, JEFFREY S.; MAHN, JEFFREY A.; BERRY, DONALD T.; SCHWERS, NORMAN F.; VANDERBEEK, THOMAS E.; NAEGELI, ROBERT E.

    2000-01-01

    This Safety Analysis Report (SAR) is prepared in compliance with the requirements of DOE Order 5480.23, Nuclear Safety Analysis Reports, and has been written to the format and content guide of DOE-STD-3009-94 Preparation Guide for U. S. Department of Energy Nonreactor Nuclear Safety Analysis Reports. The Hot Cell Facility is a Hazard Category 2 nonreactor nuclear facility, and is operated by Sandia National Laboratories for the Department of Energy. This SAR provides a description of the HCF and its operations, an assessment of the hazards and potential accidents which may occur in the facility. The potential consequences and likelihood of these accidents are analyzed and described. Using the process and criteria described in DOE-STD-3009-94, safety-related structures, systems and components are identified, and the important safety functions of each SSC are described. Additionally, information which describes the safety management programs at SNL are described in ancillary chapters of the SAR

  12. Manpower analysis in transportation safety. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Bauer, C.S.; Bowden, H.M.; Colford, C.A.; DeFilipps, P.J.; Dennis, J.D.; Ehlert, A.K.; Popkin, H.A.; Schrader, G.F.; Smith, Q.N.

    1977-05-01

    The project described provides a manpower review of national, state and local needs for safety skills, and projects future manning levels for transportation safety personnel in both the public and private sectors. Survey information revealed that there are currently approximately 121,000 persons employed directly in transportation safety occupations within the air carrier, highway and traffic safety, motor carrier, pipeline, rail carrier, and marine carrier transportation industry groups. The projected need for 1980 is over 145,000 of which over 80 percent will be in highway safety. An analysis of transportation tasks is included, and shows ten general categories about which the majority of safety activities are focused. A skills analysis shows a generally high level of educational background and several years of experience are required for most transportation safety jobs. An overall review of safety programs in the transportation industry is included, together with chapters on the individual transportation modes.

  13. Safety analysis of a high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Shimazu, Akira; Morimoto, Toshio

    1975-01-01

    In recent years, in order to satisfy the social requirements of environment and safety and also to cope with the current energy stringency, the installation of safe nuclear power plants is indispensable. Herein, safety analysis and evaluation to confirm quantitatively the safety design of a nuclear power plant become more and more important. The safety analysis and its methods for a high temperature gas-cooled reactor are described, with emphasis placed on the practices by Fuji Electric Manufacturing Co. Fundamental rule of securing plant safety ; safety analysis in normal operation regarding plant dynamic characteristics and radioactivity evaluation ; and safety analysis at the time of accidents regarding plant response to the accidents and radioactivity evaluation are explained. (Mori, K.)

  14. A new fabrication route for SFR fuel using (U, Pu)O{sub 2} powder obtained by oxalic co-conversion

    Energy Technology Data Exchange (ETDEWEB)

    Vaudez, Stéphane, E-mail: stephane.vaudez@cea.fr [CEA, DEN, DEC, SPUA, Cadarache, F-13108 Saint-Paul-Lez-Durance (France); Belin, Renaud C.; Aufore, Laurence; Sornay, Philippe [CEA, DEN, DEC, SPUA, Cadarache, F-13108 Saint-Paul-Lez-Durance (France); Grandjean, Stéphane [CEA, DEN, DRCP, DIR, Marcoule, F-30207 Bagnols sur Cèze (France)

    2013-11-15

    The standard powder metallurgy preparation of SFR (Sodium Fast Reactor) oxide fuel involves UO{sub 2} and PuO{sub 2} co-milling. An alternative route, using a solid-solution of mixed oxide obtained by oxalic co-conversion as the starting material, is presented. It was used to manufacture nuclear fuels for the “COPIX” irradiation conducted in the Phenix SFR. Two processes using co-converted powders were tested to elaborate fuel pellets: (1) the Direct Process that consists in pressing and sintering the mixed oxide with the final Pu content and (2) the Dilution Process, which involves the dilution of a high Pu content mixed oxide with UO{sub 2}. After studying the structural and microstructural evolution with temperature of these innovative raw materials, the elaboration parameters were adjusted to obtain final pellets in accordance with the Phenix fuel specifications. This study demonstrates the feasibility of such new fabrication route at laboratory scale and, from a more fundamental prospect, allows a better understanding of the underlying phenomena involved during sintering.

  15. Toward a Mechanistic Source Term in Advanced Reactors: A Review of Past U.S. SFR Incidents, Experiments, and Analyses

    Energy Technology Data Exchange (ETDEWEB)

    Bucknor, Matthew; Brunett, Acacia J.; Grabaskas, David

    2016-04-17

    In 2015, as part of a Regulatory Technology Development Plan (RTDP) effort for sodium-cooled fast reactors (SFRs), Argonne National Laboratory investigated the current state of knowledge of source term development for a metal-fueled, pool-type SFR. This paper provides a summary of past domestic metal-fueled SFR incidents and experiments and highlights information relevant to source term estimations that were gathered as part of the RTDP effort. The incidents described in this paper include fuel pin failures at the Sodium Reactor Experiment (SRE) facility in July of 1959, the Fermi I meltdown that occurred in October of 1966, and the repeated melting of a fuel element within an experimental capsule at the Experimental Breeder Reactor II (EBR-II) from November 1967 to May 1968. The experiments described in this paper include the Run-Beyond-Cladding-Breach tests that were performed at EBR-II in 1985 and a series of severe transient overpower tests conducted at the Transient Reactor Test Facility (TREAT) in the mid-1980s.

  16. Applications of probabilistic risk analysis in nuclear criticality safety design

    International Nuclear Information System (INIS)

    Chang, J.K.

    1992-01-01

    Many documents have been prepared that try to define the scope of the criticality analysis and that suggest adding probabilistic risk analysis (PRA) to the deterministic safety analysis. The report of the US Department of Energy (DOE) AL 5481.1B suggested that an accident is credible if the occurrence probability is >1 x 10 -6 /yr. The draft DOE 5480 safety analysis report suggested that safety analyses should include the application of methods such as deterministic safety analysis, risk assessment, reliability engineering, common-cause failure analysis, human reliability analysis, and human factor safety analysis techniques. The US Nuclear Regulatory Commission (NRC) report NRC SG830.110 suggested that major safety analysis methods should include but not be limited to risk assessment, reliability engineering, and human factor safety analysis. All of these suggestions have recommended including PRA in the traditional criticality analysis

  17. Blind post-test analysis of Phenix End-of-Life natural circulation test with the MARS-LMR

    International Nuclear Information System (INIS)

    Jeong, Hae Yong; Ha, Kwi Seok; Kwon, Young Min; Chang, Won Pyo; Suk, Su Dong; Lee, Kwi Lim

    2010-01-01

    KAERI is developing a system analysis code, MARS-LMR, for the application to a sodium-cooled fast reactor (SFR). This code will be used as a basic tool in the design and analysis of future SFR systems in Korea. Before wide application of a system analysis code, it is required to verify and validate the code models through analyses for appropriate experimental data or analytical results. The MARS-LMR code has been developed from MARS code which had been well verified and validated for a pressurized water reactor (PWR) system. The MARS-LMR code shares the same form of governing equations and solution schemes with MARS code, which eliminates the need of independent verification procedure. However, it is required to validate the applicability of the code to an SFR system because it adopts some dedicated heat transfer models, pressure drop models, and material properties models for a sodium system. Phenix is a medium-sized pool-type SFR successfully operated for 35 years since 1973. This reactor reached its final shutdown in February 2009. An international program of Phenix end-of-life (EOL) test was followed and some valuable information was obtained from the test, which will be useful for the validation of SFR system analysis code. In the present study, the performance of MARS-LMR code is evaluated through a blind calculation with the boundary conditions measured in the real test. The post-test analysis results are also compared with the test data generated in the test

  18. Supercritical CO2 Brayton Cycle Energy Conversion System Coupled with SFR

    International Nuclear Information System (INIS)

    Cha, Jae Eun; Kim, S. O.; Seong, S. H.; Eoh, J. H.; Lee, T. H.; Choi, S. K.; Han, J. W.; Bae, S. W.

    2008-12-01

    This report contains the description of the S-CO 2 Brayton cycle coupled to KALIMER-600 as an alternative energy conversion system. For a system development, a computer code was developed to calculate heat balance of normal operation condition. Based on the computer code, the S-CO 2 Brayton cycle energy conversion system was constructed for the KALIMER-600. Computer codes were developed to analysis for the S-CO 2 turbomachinery. Based on the design codes, the design parameters were prepared to configure the KALIMER-600 S-CO 2 turbomachinery models. A one-dimensional analysis computer code was developed to evaluate the performance of the previous PCHE heat exchangers and a design data for the typical type PCHE was produced. In parallel with the PCHE-type heat exchanger design, an airfoil shape fin PCHE heat exchanger was newly designed. The new design concept was evaluated by three-dimensional CFD analyses. Possible control schemes for power control in the KALIMER-600 S-CO 2 Brayton cycle were investigated by using the MARS code. The MMS-LMR code was also developed to analyze the transient phenomena in a SFR with a supercritical CO 2 Brayton cycle to develop the control logic. Simple power reduction and recovery event was selected and analyzed for the transient calculation. For the evaluation of Na-CO 2 boundary failure event, a computer was developed to simulate the complex thermodynamic behaviors coupled with the chemical reaction between liquid sodium and CO 2 gas. The long term behavior of a Na-CO 2 boundary failure event and its consequences which lead to a system pressure transient were evaluated

  19. Software safety hazard analysis

    International Nuclear Information System (INIS)

    Lawrence, J.D.

    1996-02-01

    Techniques for analyzing the safety and reliability of analog-based electronic protection systems that serve to mitigate hazards in process control systems have been developed over many years, and are reasonably well understood. An example is the protection system in a nuclear power plant. The extension of these techniques to systems which include digital computers is not well developed, and there is little consensus among software engineering experts and safety experts on how to analyze such systems. One possible technique is to extend hazard analysis to include digital computer-based systems. Software is frequently overlooked during system hazard analyses, but this is unacceptable when the software is in control of a potentially hazardous operation. In such cases, hazard analysis should be extended to fully cover the software. A method for performing software hazard analysis is proposed in this paper

  20. ASTRID SFR prototype steam generator design evolution related to safety and cost issues - 15236

    International Nuclear Information System (INIS)

    Woaye Hune, A.; Gerber, A.; Pirus, J.P.; Soucille, L.; Rodriguez, G.; Beauchamp, F.

    2015-01-01

    One option of the ASTRID prototype power conversion systems (PCS) is the steam cycle based on the Rankine cycle, which has been implemented in the past in France in the Phenix and Superphenix Sodium Fast Reactors, and is also being implemented in other SFRs in construction in the world (Russia, India, China). This paper is placed in the context of the Rankine cycle ameliorations and in particular on the Steam Generator (SG), developed by AREVA, from the safety point of view without significant penalty on cost. Therefore significant progresses are expected to reduce sodium/water reaction accident scenario impact as much by prevention as by mitigation. ASTRID project performed a thorough comparison of SG designs with regard to these safety requirements but also including increased sodium/water segregation (inlet and outlet of sodium implemented at the bottom of the SG), material issues and cost considerations. The design is based today on a monolithic SG with alloy 800 helical tubes but equipped with geometric specificities necessary for allowing the use of a single component per secondary loop. (authors)

  1. Infusing Reliability Techniques into Software Safety Analysis

    Science.gov (United States)

    Shi, Ying

    2015-01-01

    Software safety analysis for a large software intensive system is always a challenge. Software safety practitioners need to ensure that software related hazards are completely identified, controlled, and tracked. This paper discusses in detail how to incorporate the traditional reliability techniques into the entire software safety analysis process. In addition, this paper addresses how information can be effectively shared between the various practitioners involved in the software safety analyses. The author has successfully applied the approach to several aerospace applications. Examples are provided to illustrate the key steps of the proposed approach.

  2. Potential improvements of supercritical recompression CO2 Brayton cycle by mixing other gases for power conversion system of a SFR

    International Nuclear Information System (INIS)

    Jeong, Woo Seok; Lee, Jeong Ik; Jeong, Yong Hoon

    2011-01-01

    Highlights: → S-CO 2 cycle could be enhanced by shifting the critical point of working fluids using gas mixture. → In-house cycle code was developed to analyze supercritical Brayton cycles with gas mixture. → Gas mixture candidates were selected through a screening process: CO 2 mixing with N 2 , O 2 , He, and Ar. → CO 2 -He binary mixture shows the highest cycle efficiency increase. → Lowering the critical temperature and critical pressure of the coolant has a positive effect on the total cycle efficiency. - Abstract: A sodium-cooled fast reactor (SFR) is one of the strongest candidates for the next generation nuclear reactor. However, the conventional design of a SFR concept with an indirect Rankine cycle is subjected to a possible sodium-water reaction. To prevent any hazards from sodium-water reaction, a SFR with the Brayton cycle using Supercritical Carbon dioxide (S-CO 2 ) as the working fluid can be an alternative approach to improve the current SFR design. However, the S-CO 2 Brayton cycle is more sensitive to the critical point of working fluids than other Brayton cycles. This is because compressor work is significantly decreased slightly above the critical point due to high density of CO 2 near the boundary between the supercritical state and the subcritical state. For this reason, the minimum temperature and pressure of cycle are just above the CO 2 critical point. In other words, the critical point acts as a limitation of the lowest operating condition of the cycle. In general, lowering the rejection temperature of a thermodynamic cycle can increase the efficiency. Therefore, changing the critical point of CO 2 can result in an improvement of the total cycle efficiency with the same cycle layout. A small amount of other gases can be added in order to change the critical point of CO 2 . The direction and range of the critical point variation of CO 2 depends on the mixed component and its amount. Several gases that show chemical stability with

  3. Incorporation of advanced accident analysis methodology into safety analysis reports

    International Nuclear Information System (INIS)

    2003-05-01

    The IAEA Safety Guide on Safety Assessment and Verification defines that the aim of the safety analysis should be by means of appropriate analytical tools to establish and confirm the design basis for the items important to safety, and to ensure that the overall plant design is capable of meeting the prescribed and acceptable limits for radiation doses and releases for each plant condition category. Practical guidance on how to perform accident analyses of nuclear power plants (NPPs) is provided by the IAEA Safety Report on Accident Analysis for Nuclear Power Plants. The safety analyses are performed both in the form of deterministic and probabilistic analyses for NPPs. It is customary to refer to deterministic safety analyses as accident analyses. This report discusses the aspects of using the advanced accident analysis methods to carry out accident analyses in order to introduce them into the Safety Analysis Reports (SARs). In relation to the SAR, purposes of deterministic safety analysis can be further specified as (1) to demonstrate compliance with specific regulatory acceptance criteria; (2) to complement other analyses and evaluations in defining a complete set of design and operating requirements; (3) to identify and quantify limiting safety system set points and limiting conditions for operation to be used in the NPP limits and conditions; (4) to justify appropriateness of the technical solutions employed in the fulfillment of predetermined safety requirements. The essential parts of accident analyses are performed by applying sophisticated computer code packages, which have been specifically developed for this purpose. These code packages include mainly thermal-hydraulic system codes and reactor dynamics codes meant for the transient and accident analyses. There are also specific codes such as those for the containment thermal-hydraulics, for the radiological consequences and for severe accident analyses. In some cases, codes of a more general nature such

  4. Software safety analysis practice in installation phase

    Energy Technology Data Exchange (ETDEWEB)

    Huang, H. W.; Chen, M. H.; Shyu, S. S., E-mail: hwhwang@iner.gov.t [Institute of Nuclear Energy Research, No. 1000 Wenhua Road, Chiaan Village, Longtan Township, 32546 Taoyuan County, Taiwan (China)

    2010-10-15

    This work performed a software safety analysis in the installation phase of the Lung men nuclear power plant in Taiwan, under the cooperation of Institute of Nuclear Energy Research and Tpc. The US Nuclear Regulatory Commission requests licensee to perform software safety analysis and software verification and validation in each phase of software development life cycle with Branch Technical Position 7-14. In this work, 37 safety grade digital instrumentation and control systems were analyzed by failure mode and effects analysis, which is suggested by IEEE standard 7-4.3.2-2003. During the installation phase, skew tests for safety grade network and point to point tests were performed. The failure mode and effects analysis showed all the single failure modes can be resolved by the redundant means. Most of the common mode failures can be resolved by operator manual actions. (Author)

  5. Software safety analysis practice in installation phase

    International Nuclear Information System (INIS)

    Huang, H. W.; Chen, M. H.; Shyu, S. S.

    2010-10-01

    This work performed a software safety analysis in the installation phase of the Lung men nuclear power plant in Taiwan, under the cooperation of Institute of Nuclear Energy Research and Tpc. The US Nuclear Regulatory Commission requests licensee to perform software safety analysis and software verification and validation in each phase of software development life cycle with Branch Technical Position 7-14. In this work, 37 safety grade digital instrumentation and control systems were analyzed by failure mode and effects analysis, which is suggested by IEEE standard 7-4.3.2-2003. During the installation phase, skew tests for safety grade network and point to point tests were performed. The failure mode and effects analysis showed all the single failure modes can be resolved by the redundant means. Most of the common mode failures can be resolved by operator manual actions. (Author)

  6. Guidelines for nuclear reactor equipments safety-analysis

    International Nuclear Information System (INIS)

    1978-01-01

    The safety analysis in approving the applications for nuclear reactor constructions (or alterations) is performed by the Committee on Examination of Reactor Safety in accordance with various guidelines prescribed by the Atomic Energy Commission. In addition, the above Committee set forth its own regulations for the safety analysis on common problems among various types of nuclear reactors. This book has collected and edited those guidelines and regulations. It has two parts: Part I includes the guidelines issued to date by the Atomic Energy Commission: and Part II - regulations of the Committee. Part I has collected 8 categories of guidelines which relate to following matters: nuclear reactor sites analysis guidelines and standards for their applications; standard exposure dose of plutonium; nuclear ship operation guidelines; safety design analysis guidelines for light-water type, electricity generating nuclear reactor equipments; safety evaluation guidelines for emergency reactor core cooling system of light-water type power reactors; guidelines for exposure dose target values around light-water type electricity generating nuclear reactor equipments, and guidelines for evaluation of above target values; and meteorological guidelines for the safety analysis of electricity generating nuclear reactor equipments. Part II includes regulations of the Committee concerning - the fuel assembly used in boiling-water type and in pressurized-water type reactors; techniques of reactor core heat designs, etc. in boiling-water reactors; and others

  7. Preliminary Integrated Safety Analysis Status Report

    International Nuclear Information System (INIS)

    Gwyn, D.

    2001-01-01

    This report provides the status of the potential Monitored Geologic Repository (MGR) Integrated Safety Analysis (EA) by identifying the initial work scope scheduled for completion during the ISA development period, the schedules associated with the tasks identified, safety analysis issues encountered, and a summary of accomplishments during the reporting period. This status covers the period from October 1, 2000 through March 30, 2001

  8. Swi5-Sfr1 protein stimulates Rad51-mediated DNA strand exchange reaction through organization of DNA bases in the presynaptic filament.

    KAUST Repository

    Fornander, Louise H; Renodon-Corniè re, Axelle; Kuwabara, Naoyuki; Ito, Kentaro; Tsutsui, Yasuhiro; Shimizu, Toshiyuki; Iwasaki, Hiroshi; Nordé n, Bengt; Takahashi, Masayuki

    2013-01-01

    The Swi5-Sfr1 heterodimer protein stimulates the Rad51-promoted DNA strand exchange reaction, a crucial step in homologous recombination. To clarify how this accessory protein acts on the strand exchange reaction, we have analyzed how the structure

  9. Core Design Concept and Core Structural Material Development for a Prototype SFR

    International Nuclear Information System (INIS)

    Chang, Jinwook

    2013-01-01

    • Core design Concept: – Initial core is Uranium metal fueled core, then it will evolve into TRU core; – Tight pressure drop constraint lowers power density; – Trade-off studies with relaxed pressure drop constraint (~0.4MPa) are on-going; – Major feature will be finalized this year. • KAERI is developing advanced cladding for high burnup fuel in Ptototype SFR: – Advanced cladding materials are now developing, which shows superior high temperature mechanical property to the conventional material; – Processing technologies related to tube making process are now developed to enhance high temperature mechanical propertyl – Preliminary HT9 cladding tube was manufactured and out-of pile mechanical properties were evaluated. Advanced cladding tube is now being developed and being prepared for irradiation test

  10. Development of the On-line Acoustic Leak Detection Tool for the SFR Steam Generator Protection

    International Nuclear Information System (INIS)

    Kim, Tae-Joon; Jeong, Ji-Young; Kim, Jong-Man; Kim, Byung-Ho; Kim, Seong-O

    2007-01-01

    The successful detection of a water/steam into a sodium leak in the SFR SG (steam generator) at an early phase of a leak origin depends on the fast response and sensitivity of a leak detection system. This intention of an acoustic leak detection system is stipulated by a key impossibility of a fast detecting of an intermediate leak by the present nominal systems such as the hydrogen meter. Subject of this study is to introduce the detection performance of an on-line acoustic leak detection tool discriminated by a back-propagation neural network with a preprocessing of the 1/m Octave band analysis, and to introduce the status of an on-line development being developed with the acoustic leak detection tool(S/W) in KAERI. For a performance test, it was used with the acoustic signals for a sodium-water reaction from the injected steam into water experiments in KAERI, the acoustic signals injected from the water into the sodium obtained in IPPE, and the background noise of the PFR superheater

  11. A formal safety analysis for PLC software-based safety critical system using Z

    International Nuclear Information System (INIS)

    Koh, Jung Soo; Seong, Poong Hyun

    1997-01-01

    This paper describes a formal safety analysis technique which is demonstrated by performing empirical formal safety analysis with the case study of beamline hutch door Interlock system that is developed by using PLC (Programmable Logic Controller) systems at the Pohang Accelerator Laboratory. In order to perform formed safety analysis, we have built the Z formal specifications representation from user requirement written in ambiguous natural language and target PLC ladder logic, respectively. We have also studied the effective method to express typical PLC timer component by using specific Z formal notation which is supported by temporal history. We present a formal proof technique specifying and verifying that the hazardous states are not introduced into ladder logic in the PLC-based safety critical system

  12. A formal safety analysis for PLC software-based safety critical system using Z

    International Nuclear Information System (INIS)

    Koh, Jung Soo

    1997-02-01

    This paper describes a formal safety analysis technique which is demonstrated by performing empirical formal safety analysis with the case study of beamline hutch door Interlock system that is developed by using PLC (Programmable Logic Controller) systems at the Pohang Accelerator Laboratory. In order to perform formal safety analysis, we have built the Z formal specifications representation from user requirement written in ambiguous natural language and target PLC ladder logic, respectively. We have also studied the effective method to express typical PLC timer component by using specific Z formal notation which is supported by temporal history. We present a formal proof technique specifying and verifying that the hazardous states are not introduced into ladder logic in the PLC-based safety critical system. And also, we have found that some errors or mismatches in user requirement and final implemented PLC ladder logic while analyzing the process of the consistency and completeness of Z translated formal specifications. In the case of relatively small systems like Beamline hutch door interlock system, a formal safety analysis including explicit proof is highly recommended so that the safety of PLC-based critical system may be enhanced and guaranteed. It also provides a helpful benefits enough to comprehend user requirement expressed by ambiguous natural language

  13. An overview-probabilistic safety analysis for research reactors

    International Nuclear Information System (INIS)

    Liu Jinlin; Peng Changhong

    2015-01-01

    For long-term application, Probabilistic Safety Analysis (PSA) has proved to be a valuable tool for improving the safety and reliability of power reactors. In China, 'Nuclear safety and radioactive pollution prevention 'Twelfth Five Year Plan' and the 2020 vision' raises clearly that: to develop probabilistic safety analysis and aging evaluation for research reactors. Comparing with the power reactors, it reveals some specific features in research reactors: lower operating power, lower coolant temperature and pressure, etc. However, the core configurations may be changed very often and human actions play an important safety role in research reactors due to its specific experimental requirement. As a result, there is a necessary to conduct the PSA analysis of research reactors. This paper discusses the special characteristics related to the structure and operation and the methods to develop the PSA of research reactors, including initiating event analysis, event tree analysis, fault tree analysis, dependent failure analysis, human reliability analysis and quantification as well as the experimental and external event evaluation through the investigation of various research reactors and their PSAs home and abroad, to provide the current situation and features of research reactors PSAs. (author)

  14. Ignalina Safety Analysis Group's report for the year 1998

    International Nuclear Information System (INIS)

    Uspuras, E.; Augutis, J.; Bubelis, E.; Cesna, B.; Kaliatka, A.

    1999-02-01

    Results of Ignalina NPP Safety Analysis Group's research are presented. The main fields of group's activities in 1998 were following: safety analysis of reactor's cooling system, safety analysis of accident localization system, investigation of the problem graphite - fuel channel, reactor core modelling, assistance to the regulatory body VATESI in drafting regulations and reviewing safety reports presented by Ignalina NPP during the process of licensing of unit 1

  15. Operating plant safety analysis needs

    International Nuclear Information System (INIS)

    Young, M.Y.; Love, D.S.

    1992-01-01

    The primary objective for nuclear power station owners is to operate and manage their plants safely. However, there is also a need to provide economical electric power, which requires that the unit be operated as efficiently as possible, consistent with the safety requirements. The objectives cited above can be achieved through the identification and use of available margins inherent in the plant design. As a result of conservative licensing and analytical approaches taken in the past, many of these margins may be found in the safety analysis limits within which plants currently operate. Improvements in the accuracy of the safety analysis, and a more realistic treatment of plant initial and boundary conditions, can make this margin available for a variety of uses which enhance plant performance, help to reduce O and M costs, and may help to extend licensed operation. Opportunities for improvement exist in several areas in the accident analysis normally performed for Chapter 15 of the FSAR. For example, recent modifications to the ECCS rule, 10CFR50.46 and Appendix K, allow use of margins previously unavailable in the analysis of the Loss of Coolant Accident (LOCA). To take advantage of this regulatory change, new methods are being developed to analyze both the large and small break loss of coolant accident (LOCA). As this margin is used, enhancements in the analysis of other transients will become necessary. The paper discusses accident analysis methods, future development needs, and analysis margin utilization in specific accident scenarios

  16. Staff Nurse Perceptions of Open-Pod and Single Family Room NICU Designs on Work Environment and Patient Care.

    Science.gov (United States)

    Winner-Stoltz, Regina; Lengerich, Alexander; Hench, Anna Jeanine; OʼMalley, Janet; Kjelland, Kimberly; Teal, Melissa

    2018-06-01

    Neonatal intensive care units have historically been constructed as open units or multiple-bed bays, but since the 1990s, the trend has been toward single family room (SFR) units. The SFR design has been found to promote family-centered care and to improve patient outcomes and safety. The impact of the SFR design NICU on staff, however, has been mixed. The purposes of this study were to compare staff nurse perceptions of their work environments in an open-pod versus an SFR NICU and to compare staff nurse perceptions of the impact of 2 NICU designs on the care they provide for patients/families. A prospective cohort study was conducted. Questionnaires were completed at 6 months premove and again at 3, 9, and 15 months postmove. A series of 1-way analyses of variance were conducted to compare each group in each of the 8 domains. Open-ended questions were evaluated using thematic analysis. The SFR design is favorable in relation to environmental quality and control of primary workspace, privacy and interruption, unit features supporting individual work, and unit features supporting teamwork; the open-pod design is preferable in relation to walking. Incorporating design features that decrease staff isolation and walking and ensuring both patient and staff safety and security are important considerations. Further study is needed on unit design at a microlevel including headwall design and human milk mixing areas, as well as on workflow processes.

  17. Computer codes for safety analysis

    International Nuclear Information System (INIS)

    Holland, D.F.

    1986-11-01

    Computer codes for fusion safety analysis have been under development in the United States for about a decade. This paper will discuss five codes that are currently under development by the Fusion Safety Program. The purpose and capability of each code will be presented, a sample given, followed by a discussion of the present status and future development plans

  18. Progress of thermal hydraulic evaluation methods and experimental studies on a sodium-cooled fast reactor and its safety in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Kamide, Hideki, E-mail: kamide.hideki@jaea.go.jp; Ohshima, Hiroyuki, E-mail: ohshima.hiroyuki@jaea.go.jp; Sakai, Takaaki, E-mail: sakai.takaaki@jaea.go.jp; Tanaka, Masaaki, E-mail: tanaka.masaaki@jaea.go.jp

    2017-02-15

    Highlights: • Thermal hydraulic issues for safety design criteria of sodium cooled fast reactors. • Measurement of velocity data in a subchannel surrounded by wire wrapped fuel-pins. • Statistical evaluation of core hot spot temperature during natural circulation. • Simulation of dynamics of molten fuel pool in a core disruptive accident. • V&V procedure of a multi-dimensional thermal hydraulic code on thermal striping. - Abstract: In the framework of the Generation-IV International Forum, the safety design criteria (SDC) incorporating safety-related R&D results on innovative technologies and lessons learned from Fukushima Dai-ichi nuclear power plants accident has been established to provide the set of general criteria for the safety designs of structures, systems and components of Generation-IV Sodium-cooled Fast Reactors (Gen-IV SFRs). A number of thermal-hydraulic evaluations are necessary to meet the concept of the criteria in the design studies of Gen-IV SFRs. This paper focuses on four kinds of thermal-hydraulic issues associated with the SDC, i.e., fuel subassembly thermal-hydraulics, natural circulation decay heat removal, core disruptive accidents, and thermal striping. Progress of evaluation methods on these issues is shown with activities on verification and validation (V&V) and experimental studies towards commercialization of SFR in Japan. These evaluation methods are planned to be eventually integrated into a comprehensive numerical simulation system that can be applied to all possible phenomena in SFR systems and that can be expected to become an effective tool for the development of human resource and the handing our knowledge and technologies down.

  19. Removing unreasonable conservatisms in DOE safety analysis

    International Nuclear Information System (INIS)

    BISHOP, G.E.

    1999-01-01

    While nuclear safety analyses must always be conservative, invoking excessive conservatisms does not provide additional margins of safety. Rather, beyond a fairly narrow point, conservatisms skew a facility's true safety envelope by exaggerating risks and creating unreasonable bounds on what is required for safety. The conservatism has itself become unreasonable. A thorough review of the assumptions and methodologies contained in a facility's safety analysis can provide substantial reward, reducing both construction and operational costs without compromising actual safety

  20. HANFORD SAFETY ANALYSIS and RISK ASSESSMENT HANDBOOK (SARAH)

    International Nuclear Information System (INIS)

    EVANS, C.B.

    2004-01-01

    The purpose of the Hanford Safety Analysis and Risk Assessment Handbook (SARAH) is to support the development of safety basis documentation for Hazard Category 2 and 3 (HC-2 and 3) U.S. Department of Energy (DOE) nuclear facilities to meet the requirements of 10 CFR 830, ''Nuclear Safety Management''. Subpart B, ''Safety Basis Requirements.'' Consistent with DOE-STD-3009-94, Change Notice 2, ''Preparation Guide for U.S. Department of Energy Nonreactor Nuclear Facility Documented Safety Analyses'' (STD-3009), and DOE-STD-3011-2002, ''Guidance for Preparation of Basis for Interim Operation (BIO) Documents'' (STD-3011), the Hanford SARAH describes methodology for performing a safety analysis leading to development of a Documented Safety Analysis (DSA) and derivation of Technical Safety Requirements (TSR), and provides the information necessary to ensure a consistently rigorous approach that meets DOE expectations. The DSA and TSR documents, together with the DOE-issued Safety Evaluation Report (SER), are the basic components of facility safety basis documentation. For HC-2 or 3 nuclear facilities in long-term surveillance and maintenance (S and M), for decommissioning activities, where source term has been eliminated to the point that only low-level, residual fixed contamination is present, or for environmental remediation activities outside of a facility structure, DOE-STD-1120-98, ''Integration of Environment, Safety, and Health into Facility Disposition Activities'' (STD-1120), may serve as the basis for the DSA. HC-2 and 3 environmental remediation sites also are subject to the hazard analysis methodologies of this standard

  1. Subchannel analysis of a small ultra-long cycle fast reactor core

    International Nuclear Information System (INIS)

    Seo, Han; Kim, Ji Hyun; Bang, In Cheol

    2014-01-01

    Highlights: • The UCFR-100 is small-sized one of 60 years long-life nuclear reactors without refueling. • The design safety limits of the UCFR-100 are evaluated using MATRA-LMR. • The subchannel results are below the safety limits of general SFR design criteria. - Abstract: Thermal-hydraulic evaluation of a small ultra-long cycle fast reactor (UCFR) core is performed based on existing safety regulations. The UCFR is an innovative reactor newly designed with long-life core based on the breed-and-burn strategy and has a target electric power of 100 MWe (UCFR-100). Low enriched uranium (LEU) located at the bottom region of the core play the role of igniter to operate the UCFR for 60 years without refueling. A metallic form is selected as a burning fuel region material after the LEU location. HT-9 and sodium are used as cladding and coolant materials, respectively. In the present study, MATRA-LMR, subchannel analysis code, is used for evaluating the safety design limit of the UCFR-100 in terms of fuel, cladding, and coolant temperature distributions in the core as design criteria of a general fast reactor. The start-up period (0 year of operation), the middle of operating period (30 years of operation), and the end of operating cycle (60 years of operation) are analyzed and evaluated. The maximum cladding surface temperature (MCST) at the BOC (beginning of core life) is 498 °C on average and 551 °C when considering peaking factor, while the MCST at the MOC (middle of core life) is 498 °C on average and 548 °C in the hot channel, respectively, and the MCST at the EOC (end of core life) is 499 °C on average and 538 °C in the hot channel, respectively. The maximum cladding surface temperature over the long cycle is found at the BOC due to its high peaking factor. It is found that all results including fuel rods, cladding, and coolant exit temperature are below the safety limit of general SFR design criteria

  2. Safety assessment of research reactors and preparation of the safety analysis report

    International Nuclear Information System (INIS)

    1994-01-01

    This Safety Guide presents guidelines, approved by international consensus, for the preparation, review and assessment of safety documentation for research reactors such as the Safety Analysis Report. While the Guide is most applicable to research reactors in the design and construction stage, it is also recommended for use during relicensing or reassessment of existing reactors

  3. Application of Software Safety Analysis Methods

    International Nuclear Information System (INIS)

    Park, G. Y.; Hur, S.; Cheon, S. W.; Kim, D. H.; Lee, D. Y.; Kwon, K. C.; Lee, S. J.; Koo, Y. H.

    2009-01-01

    A fully digitalized reactor protection system, which is called the IDiPS-RPS, was developed through the KNICS project. The IDiPS-RPS has four redundant and separated channels. Each channel is mainly composed of a group of bistable processors which redundantly compare process variables with their corresponding setpoints and a group of coincidence processors that generate a final trip signal when a trip condition is satisfied. Each channel also contains a test processor called the ATIP and a display and command processor called the COM. All the functions were implemented in software. During the development of the safety software, various software safety analysis methods were applied, in parallel to the verification and validation (V and V) activities, along the software development life cycle. The software safety analysis methods employed were the software hazard and operability (Software HAZOP) study, the software fault tree analysis (Software FTA), and the software failure modes and effects analysis (Software FMEA)

  4. Nuclear safety in Slovak Republic. Safety analysis reports for WWER 440 reactors

    International Nuclear Information System (INIS)

    Rohar, S.

    1999-01-01

    Implementation of nuclear power program is connected to establishment of regulatory body for safe regulation of siting, construction, operation and decommissioning of nuclear installations. Licensing being one of the most important regulatory surveillance activity is based on independent regulatory review and assessment of information on nuclear safety for particular nuclear facility. Documents required to be submitted to the regulatory body by the licensee in Slovakia for the review and assessment usually named Safety Analysis Report (SAR) are presented in detail in this paper. Current status of Safety Analysis Reports for Bohunice V-1, Bohunice V-2 and Mochovce NPP is shown

  5. From Safety Analysis to Formal Specification

    DEFF Research Database (Denmark)

    Hansen, Kirsten Mark; Ravn, Anders P.; Stavridou, Victoria

    1998-01-01

    Software for safety critical systems must deal with the hazards identified bysafety analysis. This paper investigates, how the results of onesafety analysis technique, fault trees, are interpreted as software safetyrequirements to be used in the program design process. We propose thatfault tree...... analysis and program development use the samesystem model. This model is formalized in areal-time, interval logic, based on a conventional dynamic systems modelwith state evolving over time. Fault trees are interpreted astemporal formulas, and it is shown how such formulas can be usedfor deriving safety...

  6. Business of Nuclear Safety Analysis Office, Nuclear Technology Test Center

    International Nuclear Information System (INIS)

    Hayakawa, Masahiko

    1981-01-01

    The Nuclear Technology Test Center established the Nuclear Safety Analysis Office to execute newly the works concerning nuclear safety analysis in addition to the works related to the proving tests of nuclear machinery and equipments. The regulations for the Nuclear Safety Analysis Office concerning its organization, business and others were specially decided, and it started the business formally in August, 1980. It is a most important subject to secure the safety of nuclear facilities in nuclear fuel cycle as the premise of developing atomic energy. In Japan, the strict regulation of safety is executed by the government at each stage of the installation, construction, operation and maintenance of nuclear facilities, based on the responsibility for the security of installers themselves. The Nuclear Safety Analysis Office was established as the special organ to help the safety examination related to the installation of nuclear power stations and others by the government. It improves and puts in order the safety analysis codes required for the cross checking in the safety examination, and carries out safety analysis calculation. It is operated by the cooperation of the Science and Technology Agency and the Agency of Natural Resources and Energy. The purpose of establishment, the operation and the business of the Nuclear Safety Analysis Office, the plan of improving and putting in order of analysis codes, and the state of the similar organs in foreign countries are described. (Kako, I.)

  7. Experimental and Numerical Analysis of S-CO{sub 2} Critical Flow for SFR Recovery System Design

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Min Seok; Jung, Hwa-Young; Ahn, Yoonhan; Lee, Jekyoung; Lee, Jeong Ik [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2016-05-15

    This paper presents both numerical and experimental studies of the critical flow of S-CO{sub 2} while special attention is given to the turbo-machinery seal design. A computational critical flow model is described first. The experiments were conducted to validate the critical flow model. Various conditions have been tested to study the flow characteristic and provide validation data for the model. The comparison of numerical and experimental results of S-CO{sub 2} critical flow will be presented. In order to eliminate SWR, a concept of coupling the Supercritical CO{sub 2} (S-CO{sub 2}) cycle with SFR has been proposed. It is known that for a closed system controlling the inventory is important for stable operation and achieving high efficiency. Since the S-CO{sub 2} power cycle is a highly pressurized system, certain amount of leakage flow is inevitable in the rotating turbo-machinery via seals. To simulate the CO{sub 2} leak flow in a turbo-machinery with higher accuracy in the future, the real gas effect and friction factor will be considered for the CO{sub 2} critical flow model. Moreover, experimentally obtained temperature data were somewhat different from the numerically obtained temperature due to the insufficient insulation and large thermal inertia of the CO{sub 2} critical flow facility. Insulation in connecting pipes and the low-pressure tank will be added and additional tests will be conducted.

  8. Risk perception, safety goals and regulatory decision-making

    International Nuclear Information System (INIS)

    Hoegberg, Lars

    1998-01-01

    Deciding on 'how safe is safe enough?' includes value judgements with implications of an ethical and political nature. As regulators are accountable to governments, parliaments and the general public, regulatory decision-making should be characterized by transparency with respect to how such value judgements are reflected in risk assessments and regulatory decisions. Some approaches in this respect are discussed in the paper, based on more than fifteen years of experience in nuclear regulatory decision-making. Issues discussed include: (1) risk profiles and safety goals associated with severe reactor accidents--individual health risks, societal risks and risk of losing investments; (2) risk profile-based licensing of the Swedish SFR final disposal facility for low and intermediate level radioactive waste

  9. Establishment of Safety Analysis System and Technology for CANDU Reactors

    International Nuclear Information System (INIS)

    Park, Joo Hwan; Rhee, B. W.; Min, B. J.; Kim, H. T.; Kim, W. Y.; Yoon, C.; Chun, J. S.; Cho, M. S.; Jeong, J. Y.; Kang, H. S.

    2007-06-01

    The following 4 research items have been studied to establish a CANDU safety analysis system and to develop the relevant elementary technology for CANDU reactors. First, to improve and validate the CANDU design and operational safety analysis codes, the CANDU physics cell code WIMS-CANDU was improved, and validated, and an analysis of the moderator subcooling and pressure tube integrity has been performed for the large break LOCAs without ECCS. Also a CATHENA model and a CFD model for a post-blowdown fuel channel analysis have been developed and validated against two high temperature thermal-chemical experiments, CS28-1 and 2. Second, to improve the integrated operating system of the CANDU safety analysis codes, an extension has been made to them to include the core and fuel accident analyses, and a web-based CANDU database, CANTHIS version 2.0 was completed. Third, to assess the applicability of the ACR-7 safety analysis methodology to CANDU-6 the ACR-7 safety analysis methods were reviewed and the safety analysis methods of ACR-7 applicable to CANDU-6 were recommended. Last, to supplement and improve the existing CANDU safety analysis procedures, detailed analysis procedures have been prepared for individual accident scenarios. The results of this study can be used to resolve the CANDU safety issues, to improve the current design and operational safety analysis codes, and to technically support the Wolsong site to resolve their problems

  10. Establishment of Safety Analysis System and Technology for CANDU Reactors

    International Nuclear Information System (INIS)

    Min, Byung Joo; Kim, W. Y.; Kim, H. T.; Rhee, B. W.; Yoon, C.; Kang, H. S.; Yoo, K. J.

    2005-03-01

    To improve the CANDU design/operation safety analysis codes and the CANDU safety analysis methodology, the following works have been done. From the development of the lattice codes (WIMS/CANDU), the lattice model simulates the real core lattice geometry and the effect of the pressure tube creep to the core lattice parameter has been evaluated. From the development of the 3-dimensional thermal-hydraulic analysis model of the moderator behavior (CFX4-CAMO), validation of the model against STERN Lab experiment has been executed. The butterfly-shaped grid structure and the 3-dimensional flow resistance model for porous media were developed and applied to the moderator analysis for Wolsong units 2/3/4. The single fuel channel analysis codes for blowdown and post-blowdown were unified by CATHENA. The 3-dimensional fuel channel analysis model (CFX-CACH) has been developed for validation of CATHENA fuel channel analysis model. The interlinking analysis system (CANVAS) of the thermal-hydraulic safety analysis codes for the primary heat transport system and containment system has been executed. The database system of core physics and thermal-hydraulics experimental data for safety analysis has been established on the URL: http://CANTHIS.kaeri.re.kr. For documentation and Standardization of the general safety analysis procedure, the general safety analysis procedure is developed and applied to a large break LOCA. The present research results can be utilized for establishment of the independent safety analysis technology and acquisition of the optimal safety analysis technology

  11. Ignalina NPP Safety Analysis: Models and Results

    International Nuclear Information System (INIS)

    Uspuras, E.

    1999-01-01

    Research directions, linked to safety assessment of the Ignalina NPP, of the scientific safety analysis group are presented: Thermal-hydraulic analysis of accidents and operational transients; Thermal-hydraulic assessment of Ignalina NPP Accident Localization System and other compartments; Structural analysis of plant components, piping and other parts of Main Circulation Circuit; Assessment of RBMK-1500 reactor core and other. Models and main works carried out last year are described. (author)

  12. Reload safety analysis automation tools

    International Nuclear Information System (INIS)

    Havlůj, F.; Hejzlar, J.; Vočka, R.

    2013-01-01

    Performing core physics calculations for the sake of reload safety analysis is a very demanding and time consuming process. This process generally begins with the preparation of libraries for the core physics code using a lattice code. The next step involves creating a very large set of calculations with the core physics code. Lastly, the results of the calculations must be interpreted, correctly applying uncertainties and checking whether applicable limits are satisfied. Such a procedure requires three specialized experts. One must understand the lattice code in order to correctly calculate and interpret its results. The next expert must have a good understanding of the physics code in order to create libraries from the lattice code results and to correctly define all the calculations involved. The third expert must have a deep knowledge of the power plant and the reload safety analysis procedure in order to verify, that all the necessary calculations were performed. Such a procedure involves many steps and is very time consuming. At ÚJV Řež, a.s., we have developed a set of tools which can be used to automate and simplify the whole process of performing reload safety analysis. Our application QUADRIGA automates lattice code calculations for library preparation. It removes user interaction with the lattice code and reduces his task to defining fuel pin types, enrichments, assembly maps and operational parameters all through a very nice and user-friendly GUI. The second part in reload safety analysis calculations is done by CycleKit, a code which is linked with our core physics code ANDREA. Through CycleKit large sets of calculations with complicated interdependencies can be performed using simple and convenient notation. CycleKit automates the interaction with ANDREA, organizes all the calculations, collects the results, performs limit verification and displays the output in clickable html format. Using this set of tools for reload safety analysis simplifies

  13. Software safety analysis application in installation phase

    International Nuclear Information System (INIS)

    Huang, H. W.; Yih, S.; Wang, L. H.; Liao, B. C.; Lin, J. M.; Kao, T. M.

    2010-01-01

    This work performed a software safety analysis (SSA) in the installation phase of the Lungmen nuclear power plant (LMNPP) in Taiwan, under the cooperation of INER and TPC. The US Nuclear Regulatory Commission (USNRC) requests licensee to perform software safety analysis (SSA) and software verification and validation (SV and V) in each phase of software development life cycle with Branch Technical Position (BTP) 7-14. In this work, 37 safety grade digital instrumentation and control (I and C) systems were analyzed by Failure Mode and Effects Analysis (FMEA), which is suggested by IEEE Standard 7-4.3.2-2003. During the installation phase, skew tests for safety grade network and point to point tests were performed. The FMEA showed all the single failure modes can be resolved by the redundant means. Most of the common mode failures can be resolved by operator manual actions. (authors)

  14. Software safety analysis techniques for developing safety critical software in the digital protection system of the LMR

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jang Soo; Cheon, Se Woo; Kim, Chang Hoi; Sim, Yun Sub

    2001-02-01

    This report has described the software safety analysis techniques and the engineering guidelines for developing safety critical software to identify the state of the art in this field and to give the software safety engineer a trail map between the code and standards layer and the design methodology and documents layer. We have surveyed the management aspects of software safety activities during the software lifecycle in order to improve the safety. After identifying the conventional safety analysis techniques for systems, we have surveyed in details the software safety analysis techniques, software FMEA(Failure Mode and Effects Analysis), software HAZOP(Hazard and Operability Analysis), and software FTA(Fault Tree Analysis). We have also surveyed the state of the art in the software reliability assessment techniques. The most important results from the reliability techniques are not the specific probability numbers generated, but the insights into the risk importance of software features. To defend against potential common-mode failures, high quality, defense-in-depth, and diversity are considered to be key elements in digital I and C system design. To minimize the possibility of CMFs and thus increase the plant reliability, we have provided D-in-D and D analysis guidelines.

  15. Software safety analysis techniques for developing safety critical software in the digital protection system of the LMR

    International Nuclear Information System (INIS)

    Lee, Jang Soo; Cheon, Se Woo; Kim, Chang Hoi; Sim, Yun Sub

    2001-02-01

    This report has described the software safety analysis techniques and the engineering guidelines for developing safety critical software to identify the state of the art in this field and to give the software safety engineer a trail map between the code and standards layer and the design methodology and documents layer. We have surveyed the management aspects of software safety activities during the software lifecycle in order to improve the safety. After identifying the conventional safety analysis techniques for systems, we have surveyed in details the software safety analysis techniques, software FMEA(Failure Mode and Effects Analysis), software HAZOP(Hazard and Operability Analysis), and software FTA(Fault Tree Analysis). We have also surveyed the state of the art in the software reliability assessment techniques. The most important results from the reliability techniques are not the specific probability numbers generated, but the insights into the risk importance of software features. To defend against potential common-mode failures, high quality, defense-in-depth, and diversity are considered to be key elements in digital I and C system design. To minimize the possibility of CMFs and thus increase the plant reliability, we have provided D-in-D and D analysis guidelines

  16. Status of SPACE Safety Analysis Code Development

    International Nuclear Information System (INIS)

    Lee, Dong Hyuk; Yang, Chang Keun; Kim, Se Yun; Ha, Sang Jun

    2009-01-01

    In 2006, the Korean the Korean nuclear industry started developing a thermal-hydraulic analysis code for safety analysis of PWR(Pressurized Water Reactor). The new code is named as SPACE(Safety and Performance Analysis Code for Nuclear Power Plant). The SPACE code can solve two-fluid, three-field governing equations in one dimensional or three dimensional geometry. The SPACE code has many component models required for modeling a PWR, such as reactor coolant pump, safety injection tank, etc. The programming language used in the new code is C++, for new generation of engineers who are more comfortable with C/C++ than old FORTRAN language. This paper describes general characteristics of SPACE code and current status of SPACE code development

  17. Systems Analysis of NASA Aviation Safety Program: Final Report

    Science.gov (United States)

    Jones, Sharon M.; Reveley, Mary S.; Withrow, Colleen A.; Evans, Joni K.; Barr, Lawrence; Leone, Karen

    2013-01-01

    A three-month study (February to April 2010) of the NASA Aviation Safety (AvSafe) program was conducted. This study comprised three components: (1) a statistical analysis of currently available civilian subsonic aircraft data from the National Transportation Safety Board (NTSB), the Federal Aviation Administration (FAA), and the Aviation Safety Information Analysis and Sharing (ASIAS) system to identify any significant or overlooked aviation safety issues; (2) a high-level qualitative identification of future safety risks, with an assessment of the potential impact of the NASA AvSafe research on the National Airspace System (NAS) based on these risks; and (3) a detailed, top-down analysis of the NASA AvSafe program using an established and peer-reviewed systems analysis methodology. The statistical analysis identified the top aviation "tall poles" based on NTSB accident and FAA incident data from 1997 to 2006. A separate examination of medical helicopter accidents in the United States was also conducted. Multiple external sources were used to develop a compilation of ten "tall poles" in future safety issues/risks. The top-down analysis of the AvSafe was conducted by using a modification of the Gibson methodology. Of the 17 challenging safety issues that were identified, 11 were directly addressed by the AvSafe program research portfolio.

  18. Status of generic actions items and safety analysis system of PHWR

    Energy Technology Data Exchange (ETDEWEB)

    Park, Joo Hwan; Min, Byung Joo

    2001-05-01

    This report described the review results of a GAIs(Generic Action Item) currently issued on safety analysis of PHWR(Pressurized Heavy Water Reactor) and the research activities and positions to solve the GAIs in each country which possess PHWRs. eviewing the Final Safety Analysis Report for Wolsong-2/3/4 Units, the safety analysis methodology, classification for accident scenarios, safety analysis codes, their interface, etc.. were described. From the present review report, it is intended to establish the CANDU safety analysis system by providing the better understandings and development plans for the safety analysis of PHWR. esults.

  19. 2005 dossier: granite. Tome: safety analysis of the geologic disposal

    International Nuclear Information System (INIS)

    2005-01-01

    This document makes a status of the researches carried out by the French national agency of radioactive wastes (ANDRA) about the safety aspects of the geologic disposal of high-level and long-lived (HLLL) radioactive wastes in granite formations. Content: 1 - safety approach: context and general goal, references, design approach by safety functions, safety approach during the construction-exploitation-observation-closure phase, safety analysis during the post-closure phase; 2 - general description: HLLL wastes, granitic environment, general structure of the architecture of a disposal facility; 3 - safety functions and disposal design: general context, safety functions of the long-term disposal, design dispositions retained to answer the functions; 4 - operational safety: people's protection, radiological risks during exploitation, risk analysis in accident situation; 5 - qualitative safety analysis: methodology, main results of the analysis of the features, events and processes (FEP) database; 6 - disposal efficiency evaluation during post-closure phase: calculation models, calculation tools used for the modeling of radionuclides transport, calculation results and main lessons. (J.S.)

  20. Integrated framework for dynamic safety analysis

    International Nuclear Information System (INIS)

    Kim, Tae Wan; Karanki, Durga R.

    2012-01-01

    In the conventional PSA (Probabilistic Safety Assessment), detailed plant simulations by independent thermal hydraulic (TH) codes are used in the development of accident sequence models. Typical accidents in a NPP involve complex interactions among process, safety systems, and operator actions. As independent TH codes do not have the models of operator actions and full safety systems, they cannot literally simulate the integrated and dynamic interactions of process, safety systems, and operator responses. Offline simulation with pre decided states and time delays may not model the accident sequences properly. Moreover, when stochastic variability in responses of accident models is considered, defining all the combinations for simulations will be cumbersome task. To overcome some of these limitations of conventional safety analysis approach, TH models are coupled with the stochastic models in the dynamic event tree (DET) framework, which provides flexibility to model the integrated response due to better communication as all the accident elements are in the same model. The advantages of this framework also include: Realistic modeling in dynamic scenarios, comprehensive results, integrated approach (both deterministic and probabilistic models), and support for HRA (Human Reliability Analysis)

  1. OASIS: An automotive analysis and safety engineering instrument

    International Nuclear Information System (INIS)

    Mader, Roland; Armengaud, Eric; Grießnig, Gerhard; Kreiner, Christian; Steger, Christian; Weiß, Reinhold

    2013-01-01

    In this paper, we describe a novel software tool named OASIS (AutOmotive Analysis and Safety EngIneering InStrument). OASIS supports automotive safety engineering with features allowing the creation of consistent and complete work products and to simplify and automate workflow steps from early analysis through system development to software development. More precisely, it provides support for (a) model creation and reuse, (b) analysis and documentation and (c) configuration and code generation. We present OASIS as a part of a tool chain supporting the application of a safety engineering workflow aligned with the automotive safety standard ISO 26262. In particular, we focus on OASIS' (1) support for property checking and model correction as well as its (2) support for fault tree generation and FMEA (Failure Modes and Effects Analysis) table generation. Finally, based on the case study of hybrid electric vehicle development, we demonstrate that (1) and (2) are able to strongly support FTA (Fault Tree Analysis) and FMEA

  2. PWR core safety analysis with 3-dimensional methods

    International Nuclear Information System (INIS)

    Gensler, A.; Kühnel, K.; Kuch, S.

    2015-01-01

    Highlights: • An overview of AREVA’s safety analysis codes their coupling is provided. • The validation base and licensing applications of these codes are summarized. • Coupled codes and methods provide improved margins and non-conservative results. • Examples for REA and inadvertent opening of the pressurizer safety valve are given. - Abstract: The main focus of safety analysis is to demonstrate the required safety level of the reactor core. Because of the demanding requirements, the quality of the safety analysis strongly affects the confidence in the operational safety of a reactor. To ensure the highest quality, it is essential that the methodology consists of appropriate analysis tools, an extensive validation base, and last but not least highly educated engineers applying the methodology. The sophisticated 3-dimensional core models applied by AREVA ensure that all physical effects relevant for safety are treated and the results are reliable and conservative. Presently AREVA employs SCIENCE, CASMO/NEMO and CASCADE-3D for pressurized water reactors. These codes are currently being consolidated into the next generation 3D code system ARCADIA®. AREVA continuously extends the validation base, including measurement campaigns in test facilities and comparisons of the predictions of steady state and transient measured data gathered from plants during many years of operation. Thus, the core models provide reliable and comprehensive results for a wide range of applications. For the application of these powerful tools, AREVA is taking benefit of its interdisciplinary know-how and international teamwork. Experienced engineers of different technical backgrounds are working together to ensure an appropriate interpretation of the calculation results, uncertainty analysis, along with continuously maintaining and enhancing the quality of the analysis methodologies. In this paper, an overview of AREVA’s broad application experience as well as the broad validation

  3. Safety analysis of the UTSI-CFFF superconducting magnet

    International Nuclear Information System (INIS)

    Turner, L.R.; Wang, S.T.; Smith, R.P.; VanderArend, P.C.; Hsu, Y.H.

    1979-01-01

    In designing a large superconducting magnet such as the UTSI-CFFF dipole, great attention must be devoted to the safety of the magnet and personnel. The conductor for the UTSI-CFFF magnet incorporates much copper stabilizer, which both insures its cryostability, and contributes to the magnet safety. The quench analysis and the cryostat fault condition analysis are presented. Two analyses of exposed turns follow; the first shows that gas cooling protects uncovered turns; the second, that the cryostat pressure relief system protects them. Finally the failure mode and safety analysis is presented

  4. MARS-LMR modeling for the post-test analysis of Phenix End-of-Life natural circulation

    International Nuclear Information System (INIS)

    Jeong, Hae Yong; Ha, Kwi Seok; Chang, Won Pyo; Lee, Kwi Lim

    2011-01-01

    For a successful design and analysis of Sodium cooled Fast Reactor (SFR), it is required to have a reliable and well-proven system analysis code. To achieve this purpose, KAERI is enhancing the modeling capability of MARS code by adding the SFR-specific models such as pressure drop model, heat transfer model and reactivity feedback model. This version of MARS-LMR will be used as a basic tool in the design and analysis of future SFR systems in Korea. Before wide application of MARS-LMR code, it is required to verify and validate the code models through analyses for appropriate experimental data or analytical results. The end-of-life test of Phenix reactor performed by the CEA provided a unique opportunity to have reliable test data which is very valuable in the validation and verification of a SFR system analysis code. The KAERI joined this international program of the analysis of Phenix end-of-life natural circulation test coordinated by the IAEA from 2008. The main test of natural circulation was completed in 2009. Before the test the KAERI performed the pre-test analysis based on the design condition provided by the CEA. Then, the blind post-test analysis was also performed based on the test conditions measured during the test before the CEA provide the final test results. Finally, the final post-test analysis was performed recently to predict the test results as accurate as possible. This paper introduces the modeling approach of the MARS-LMR used in the final post-test analysis and summarizes the major results of the analysis

  5. Identification of Initiating Events for PGSFR

    International Nuclear Information System (INIS)

    Kim, Jintae; Jae, Moosung

    2016-01-01

    The Sodium-cooled Fast Reactor (SFR) is by far the most advanced reactor of the six Generation IV reactors. The SFR uses liquid sodium as the reactor coolant, which has superior heat transport characteristics. It also allows high power density with low coolant volume fraction and operation at low pressure. In Korea, KAERI has been developing Prototype Generation-IV Sodium-cooled Fast Reactor (PGSFR) that employs passive safety systems and inherent reactivity feedback effects. In order to prepare for the licensing, it is necessary to assess the safety of the reactor. Thus, the objective of this study is to conduct accident sequence analysis that can contribute to risk assessment. The analysis embraces identification of initiating events and accident sequences development. PGSFR is to test and demonstrate the performance of transuranic (TRU)-containing metal fuel required for a commercial SFR, and to demonstrate the TRU transmutation capability of a burner reactor as a part of an advanced fuel cycle system. Initiating events that can happen in PGSFR were identified through the MLD method. This method presents a model of a plant in terms of individual events and their combinations in a systematic and logical way. The 11 identified initiating events in this study include the events considered in the past analysis that was conducted for PRISM-150

  6. Identification of Initiating Events for PGSFR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jintae; Jae, Moosung [Hanyang University, Seoul (Korea, Republic of)

    2016-10-15

    The Sodium-cooled Fast Reactor (SFR) is by far the most advanced reactor of the six Generation IV reactors. The SFR uses liquid sodium as the reactor coolant, which has superior heat transport characteristics. It also allows high power density with low coolant volume fraction and operation at low pressure. In Korea, KAERI has been developing Prototype Generation-IV Sodium-cooled Fast Reactor (PGSFR) that employs passive safety systems and inherent reactivity feedback effects. In order to prepare for the licensing, it is necessary to assess the safety of the reactor. Thus, the objective of this study is to conduct accident sequence analysis that can contribute to risk assessment. The analysis embraces identification of initiating events and accident sequences development. PGSFR is to test and demonstrate the performance of transuranic (TRU)-containing metal fuel required for a commercial SFR, and to demonstrate the TRU transmutation capability of a burner reactor as a part of an advanced fuel cycle system. Initiating events that can happen in PGSFR were identified through the MLD method. This method presents a model of a plant in terms of individual events and their combinations in a systematic and logical way. The 11 identified initiating events in this study include the events considered in the past analysis that was conducted for PRISM-150.

  7. Reliability Analysis for Safety Grade PLC(POSAFE-Q)

    International Nuclear Information System (INIS)

    Choi, Kyung Chul; Song, Seung Whan; Park, Gang Min; Hwang, Sung Jae

    2012-01-01

    Safety Grade PLC(Programmable Logic Controller), POSAFE-Q, was developed recently in accordance with nuclear regulatory and requirements. In this paper, describe reliability analysis for digital safety grade PLC (especially POSAFE-Q). Reliability analysis scope is Prediction, Calculation of MTBF (Mean Time Between Failure), FMEA (Failure Mode Effect Analysis), PFD (Probability of Failure on Demand). (author)

  8. Status of safety analysis reports

    Energy Technology Data Exchange (ETDEWEB)

    Cserhati, A

    1999-06-01

    The safety regulation connected to both of the Atomic Acts from 1980 and 1996 requires preparation of the Preliminary Safety Analysis Report (PSAR) as well as Final SAR (FSAR). In this respect the licensing procedure for the construction and commissioning of Paks NPP did not formally deviate from the standards applied in developed countries; this is particularly true if comparison is made with the standards applied for commissioning NPPs in the second half of the seventies. By the time the overall development of internationally accepted safety standards and some existing deficiencies of earlier SAR made necessary a general reassessment of the plant safety (AGNES project). The carried out PSR for Paks-1 and 2 also added a valuable contribution to the SAR content, however a formal update of SAR is not made yet. A Hungarian nuclear authority decree from 1997 obligates the licensee to prepare and submit a major upgrade of FSAR until the mid of 2000, after finishing the PSR for Paks-3 and 4. From this date a periodic update of FSAR is required every year. The operational license renewal affects only the PSR but not the FSAR updating. The new Nuclear Safety Code outlines the contents of PSAR and FSAR, based on US NRC Reg. Guide 1. 70. Rev. 3. Hungary by now can fulfill the upgrading of SAR without major external technical or financial help. The AGNES project covered the safety analysis chapters of SAR. It was financed mainly by the country. In the project there have been involved in limited cases as performers the VTT (Finland), Belgatom (Belgium), GRS (Germany), etc., the IVO (Finland) fulfilled tasks of an independent reviewer for safety analysis. The AGNES had certain interconnection with the similar IAEA RER safety reassessment project for WWER-440/213. The PSR for Paks-1 and 2 have been carried out by the Paks staff from the resources of the plant. During the evaluation of several parts of Paks-3 and 4 PSR documentation the authority intends to use certain

  9. Status of safety analysis reports

    International Nuclear Information System (INIS)

    Cserhati, A.

    1999-01-01

    The safety regulation connected to both of the Atomic Acts from 1980 and 1996 requires preparation of the Preliminary Safety Analysis Report (PSAR) as well as Final SAR (FSAR). In this respect the licensing procedure for the construction and commissioning of Paks NPP did not formally deviate from the standards applied in developed countries; this is particularly true if comparison is made with the standards applied for commissioning NPPs in the second half of the seventies. By the time the overall development of internationally accepted safety standards and some existing deficiencies of earlier SAR made necessary a general reassessment of the plant safety (AGNES project). The carried out PSR for Paks-1 and 2 also added a valuable contribution to the SAR content, however a formal update of SAR is not made yet. A Hungarian nuclear authority decree from 1997 obligates the licensee to prepare and submit a major upgrade of FSAR until the mid of 2000, after finishing the PSR for Paks-3 and 4. From this date a periodic update of FSAR is required every year. The operational license renewal affects only the PSR but not the FSAR updating. The new Nuclear Safety Code outlines the contents of PSAR and FSAR, based on US NRC Reg. Guide 1. 70. Rev. 3. Hungary by now can fulfill the upgrading of SAR without major external technical or financial help. The AGNES project covered the safety analysis chapters of SAR. It was financed mainly by the country. In the project there have been involved in limited cases as performers the VTT (Finland), Belgatom (Belgium), GRS (Germany), etc., the IVO (Finland) fulfilled tasks of an independent reviewer for safety analysis. The AGNES had certain interconnection with the similar IAEA RER safety reassessment project for WWER-440/213. The PSR for Paks-1 and 2 have been carried out by the Paks staff from the resources of the plant. During the evaluation of several parts of Paks-3 and 4 PSR documentation the authority intends to use certain

  10. Safety analysis of the nuclear chemistry Building 151

    International Nuclear Information System (INIS)

    Kvam, D.

    1984-01-01

    This report summarizes the results of a safety analysis that was done on Building 151. The report outlines the methodology, the analysis, and the findings that led to the low hazard classification. No further safety evaluation is indicated at this time. 5 tables

  11. DESIGN PACKAGE 1E SYSTEM SAFETY ANALYSIS

    Energy Technology Data Exchange (ETDEWEB)

    M. Salem

    1995-06-23

    The purpose of this analysis is to systematically identify and evaluate hazards related to the Yucca Mountain Project Exploratory Studies Facility (ESF) Design Package 1E, Surface Facilities, (for a list of design items included in the package 1E system safety analysis see section 3). This process is an integral part of the systems engineering process; whereby safety is considered during planning, design, testing, and construction. A largely qualitative approach was used since a radiological System Safety Analysis is not required. The risk assessment in this analysis characterizes the accident scenarios associated with the Design Package 1E structures/systems/components(S/S/Cs) in terms of relative risk and includes recommendations for mitigating all identified risks. The priority for recommending and implementing mitigation control features is: (1) Incorporate measures to reduce risks and hazards into the structure/system/component design, (2) add safety devices and capabilities to the designs that reduce risk, (3) provide devices that detect and warn personnel of hazardous conditions, and (4) develop procedures and conduct training to increase worker awareness of potential hazards, on methods to reduce exposure to hazards, and on the actions required to avoid accidents or correct hazardous conditions.

  12. The influence of sodium fires on LMFBRs safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Justin, F [DSN/Centre de Fontenay-aux-Roses, Fontenay-aux-Roses (France)

    1979-03-01

    In a sodium cooled reactor, sodium fires are accidental conditions to be taken into account in safety analysis. For the various sodium categories, fire conditions, associated risks, safety analysis objectives and detailed corresponding issues are indicated, An experimental research program can be deduced from these considerations. This report covers the following: safety analysis methodology; primary sodium fires; secondary sodium fires; auxiliary sodium fires, and related experimental research programs.

  13. The influence of sodium fires on LMFBRs safety analysis

    International Nuclear Information System (INIS)

    Justin, F.

    1979-01-01

    In a sodium cooled reactor, sodium fires are accidental conditions to be taken into account in safety analysis. For the various sodium categories, fire conditions, associated risks, safety analysis objectives and detailed corresponding issues are indicated, An experimental research program can be deduced from these considerations. This report covers the following: safety analysis methodology; primary sodium fires; secondary sodium fires; auxiliary sodium fires, and related experimental research programs

  14. Meta-analysis of surgical safety checklist effects on teamwork, communication, morbidity, mortality, and safety.

    Science.gov (United States)

    Lyons, Vanessa E; Popejoy, Lori L

    2014-02-01

    The purpose of this study is to examine the effectiveness of surgical safety checklists on teamwork, communication, morbidity, mortality, and compliance with safety measures through meta-analysis. Four meta-analyses were conducted on 19 studies that met the inclusion criteria. The effect size of checklists on teamwork and communication was 1.180 (p = .003), on morbidity and mortality was 0.123 (p = .003) and 0.088 (p = .001), respectively, and on compliance with safety measures was 0.268 (p teamwork and communication, reduce morbidity and mortality, and improve compliance with safety measures. This meta-analysis is limited in its generalizability based on the limited number of studies and the inclusion of only published research. Future research is needed to examine possible moderating variables for the effects of surgical safety checklists.

  15. Reliability analysis of software based safety functions

    International Nuclear Information System (INIS)

    Pulkkinen, U.

    1993-05-01

    The methods applicable in the reliability analysis of software based safety functions are described in the report. Although the safety functions also include other components, the main emphasis in the report is on the reliability analysis of software. The check list type qualitative reliability analysis methods, such as failure mode and effects analysis (FMEA), are described, as well as the software fault tree analysis. The safety analysis based on the Petri nets is discussed. The most essential concepts and models of quantitative software reliability analysis are described. The most common software metrics and their combined use with software reliability models are discussed. The application of software reliability models in PSA is evaluated; it is observed that the recent software reliability models do not produce the estimates needed in PSA directly. As a result from the study some recommendations and conclusions are drawn. The need of formal methods in the analysis and development of software based systems, the applicability of qualitative reliability engineering methods in connection to PSA and the need to make more precise the requirements for software based systems and their analyses in the regulatory guides should be mentioned. (orig.). (46 refs., 13 figs., 1 tab.)

  16. Issues affecting advanced passive light-water reactor safety analysis

    International Nuclear Information System (INIS)

    Beelman, R.J.; Fletcher, C.D.; Modro, S.M.

    1992-01-01

    Next generation commercial reactor designs emphasize enhanced safety through improved safety system reliability and performance by means of system simplification and reliance on immutable natural forces for system operation. Simulating the performance of these safety systems will be central to analytical safety evaluation of advanced passive reactor designs. Yet the characteristically small driving forces of these safety systems pose challenging computational problems to current thermal-hydraulic systems analysis codes. Additionally, the safety systems generally interact closely with one another, requiring accurate, integrated simulation of the nuclear steam supply system, engineered safeguards and containment. Furthermore, numerical safety analysis of these advanced passive reactor designs wig necessitate simulation of long-duration, slowly-developing transients compared with current reactor designs. The composite effects of small computational inaccuracies on induced system interactions and perturbations over long periods may well lead to predicted results which are significantly different than would otherwise be expected or might actually occur. Comparisons between the engineered safety features of competing US advanced light water reactor designs and analogous present day reactor designs are examined relative to the adequacy of existing thermal-hydraulic safety codes in predicting the mechanisms of passive safety. Areas where existing codes might require modification, extension or assessment relative to passive safety designs are identified. Conclusions concerning the applicability of these codes to advanced passive light water reactor safety analysis are presented

  17. STARS software tool for analysis of reliability and safety

    International Nuclear Information System (INIS)

    Poucet, A.; Guagnini, E.

    1989-01-01

    This paper reports on the STARS (Software Tool for the Analysis of Reliability and Safety) project aims at developing an integrated set of Computer Aided Reliability Analysis tools for the various tasks involved in systems safety and reliability analysis including hazard identification, qualitative analysis, logic model construction and evaluation. The expert system technology offers the most promising perspective for developing a Computer Aided Reliability Analysis tool. Combined with graphics and analysis capabilities, it can provide a natural engineering oriented environment for computer assisted reliability and safety modelling and analysis. For hazard identification and fault tree construction, a frame/rule based expert system is used, in which the deductive (goal driven) reasoning and the heuristic, applied during manual fault tree construction, is modelled. Expert system can explain their reasoning so that the analyst can become aware of the why and the how results are being obtained. Hence, the learning aspect involved in manual reliability and safety analysis can be maintained and improved

  18. Safety analysis methodology for OPR 1000

    International Nuclear Information System (INIS)

    Hwang-Yong, Jun

    2005-01-01

    Full text: Korea Electric Power Research Institute (KEPRI) has been developing inhouse safety analysis methodology based on the delicate codes available to KEPRI to overcome the problems arising from currently used vendor oriented methodologies. For the Loss of Coolant Accident (LOCA) analysis, the KREM (KEPRI Realistic Evaluation Methodology) has been developed based on the RELAP-5 code. The methodology was approved for the Westinghouse 3-loop plants by the Korean regulatory organization and the project to extent the methodology to the Optimized Power Reactor 1000 (OPR1000) has been ongoing since 2001. Also, for the Non-LOCA analysis, the KNAP (Korea Non-LOCA Analysis Package) has been developed using the UNICORN-TM code system. To demonstrate the feasibility of these codes systems and methodologies, some typical cases of the design basis accidents mentioned in the final safety analysis report (FSAR) were analyzed. (author)

  19. Swimming pool reactor reliability and safety analysis

    International Nuclear Information System (INIS)

    Li Zhaohuan

    1997-01-01

    A reliability and safety analysis of Swimming Pool Reactor in China Institute of Atomic Energy is done by use of event/fault tree technique. The paper briefly describes the analysis model, analysis code and main results. Meanwhile it also describes the impact of unassigned operation status on safety, the estimation of effectiveness of defense tactics in maintenance against common cause failure, the effectiveness of recovering actions on the system reliability, the comparison of occurrence frequencies of the core damage by use of generic and specific data

  20. Analysis on safety production in coal mines Henan Province

    Institute of Scientific and Technical Information of China (English)

    KONG Liu-an; ZHANG Wen-yong

    2006-01-01

    Based on the rigorous situation of safety production in coal mines, the paper analyzed the statistical data of recent accidents indexes in Henan's coal mines. Using investigation and comparison analysis methods, a specified analysis on mining conditions, technical facility level, safety input and vocational quality of workers in Henan's coal mines was conducted. The result indicates that there have been existing such main safety production problems as weak safety management, low-level facilities, inadequate safety input and poor vocational quality and so on. Finally it proposes such reference solutions as to establish and perfect coal mining supervision and management system, to increase safety investment into techniques and facilities and to strengthen workers' safety education and introduction of more high-level professional talents.

  1. Preliminary safety analysis of unscrammed events for KLFR

    International Nuclear Information System (INIS)

    Kim, S.J.; Ha, G.S.

    2005-01-01

    The report presents the design features of KLFR; Safety Analysis Code; steady-state calculation results and analysis results of unscrammed events. The calculations of the steady-state and unscrammed events have been performed for the conceptual design of KLFR using SSC-K code. UTOP event results in no fuel damage and no centre-line melting. The inherent safety features are demonstrated through the analysis of ULOHS event. Although the analysis of ULOF has much uncertainties in the pump design, the analysis results show the inherent safety characteristics. 6% flow of rated flow of natural circulation is formed in the case of ULOF. In the metallic fuel rod, the cladding temperature is somewhat high due to the low heat transfer coefficient of lead. ULOHS event should be considered in design of RVACS for long-term cooling

  2. SACS2: Dynamic and Formal Safety Analysis Method for Complex Safety Critical System

    International Nuclear Information System (INIS)

    Koh, Kwang Yong; Seong, Poong Hyun

    2009-01-01

    Fault tree analysis (FTA) is one of the most widely used safety analysis technique in the development of safety critical systems. However, over the years, several drawbacks of the conventional FTA have become apparent. One major drawback is that conventional FTA uses only static gates and hence can not capture dynamic behaviors of the complex system precisely. Although several attempts such as dynamic fault tree (DFT), PANDORA, formal fault tree (FFT) and so on, have been made to overcome this problem, they can not still do absolute or actual time modeling because they adapt relative time concept and can capture only sequential behaviors of the system. Second drawback of conventional FTA is its lack of rigorous semantics. Because it is informal in nature, safety analysis results heavily depend on an analyst's ability and are error-prone. Finally reasoning process which is to check whether basic events really cause top events is done manually and hence very labor-intensive and timeconsuming for the complex systems. In this paper, we propose a new safety analysis method for complex safety critical system in qualitative manner. We introduce several temporal gates based on timed computational tree logic (TCTL) which can represent quantitative notion of time. Then, we translate the information of the fault trees into UPPAAL query language and the reasoning process is automatically done by UPPAAL which is the model checker for time critical system

  3. MSSV Modeling for Wolsong-1 Safety Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Moon, Bok Ja; Choi, Chul Jin; Kim, Seoung Rae [KEPCO EandC, Daejeon (Korea, Republic of)

    2010-10-15

    The main steam safety valves (MSSVs) are installed on the main steam line to prevent the overpressurization of the system. MSSVs are held in closed position by spring force and the valves pop open by internal force when the main steam pressure increases to open set pressure. If the overpressure condition is relieved, the valves begin to close. For the safety analysis of anticipated accident condition, the safety systems are modeled conservatively to simulate the accident condition more severe. MSSVs are also modeled conservatively for the analysis of over-pressurization accidents. In this paper, the pressure transient is analyzed at over-pressurization condition to evaluate the conservatism for MSSV models

  4. Economic consideration of nuclear safety and cost benefit analysis in nuclear safety regulation

    International Nuclear Information System (INIS)

    Choi, Y. S.; Choi, K. S.; Choi, K. W.; Song, I. J.; Park, D. K.

    2001-01-01

    For the optimization of nuclear safety regulation, understanding of economic aspects of it becomes increasingly important together with the technical approach used so far to secure nuclear safety. Relevant economic theories on private and public goods were reviewed to re-illuminate nuclear safety from the economic perspective. The characteristics of nuclear safety as a public good was reviewed and discussed in comparison with the car safety as a private safety good. It was shown that the change of social welfare resulted from the policy change induced can be calculated by the summation of compensating variation(CV) of individuals. It was shown that the value of nuclear safety could be determined in monetary term by this approach. The theoretical background and history of cost benefit analysis of nuclear safety regulation were presented and topics for future study were suggested

  5. Safety culture and accident analysis-A socio-management approach based on organizational safety social capital

    International Nuclear Information System (INIS)

    Rao, Suman

    2007-01-01

    One of the biggest challenges for organizations in today's competitive business environment is to create and preserve a self-sustaining safety culture. Typically, Key drivers of safety culture in many organizations are regulation, audits, safety training, various types of employee exhortations to comply with safety norms, etc. However, less evident factors like networking relationships and social trust amongst employees, as also extended networking relationships and social trust of organizations with external stakeholders like government, suppliers, regulators, etc., which constitute the safety social capital in the Organization-seem to also influence the sustenance of organizational safety culture. Can erosion in safety social capital cause deterioration in safety culture and contribute to accidents? If so, how does it contribute? As existing accident analysis models do not provide answers to these questions, CAMSoC (Curtailing Accidents by Managing Social Capital), an accident analysis model, is proposed. As an illustration, five accidents: Bhopal (India), Hyatt Regency (USA), Tenerife (Canary Islands), Westray (Canada) and Exxon Valdez (USA) have been analyzed using CAMSoC. This limited cross-industry analysis provides two key socio-management insights: the biggest source of motivation that causes deviant behavior leading to accidents is 'Faulty Value Systems'. The second biggest source is 'Enforceable Trust'. From a management control perspective, deterioration in safety culture and resultant accidents is more due to the 'action controls' rather than explicit 'cultural controls'. Future research directions to enhance the model's utility through layering are addressed briefly

  6. Review of SFR In-Vessel Radiological Source Term Studies

    International Nuclear Information System (INIS)

    Suk, Soo Dong; Lee, Yong Bum

    2008-10-01

    An effort has been made in this study to search for and review the literatures in public domain on the studies of the phenomena related to the release of radionuclides and aerosols to the reactor containment of the sodium fast reactor (SFR) plants (i.e., in-vessel source term), made in Japan and Europe including France, Germany and UK over the last few decades. Review work is focused on the experimental programs to investigate the phenomena related to determining the source terms, with a brief review on supporting analytical models and computer programs. In this report, the research programs conducted to investigate the CDA (core disruptive accident) bubble behavior in the sodium pool for determining 'primary' or 'instantaneous' source term are first introduced. The studies performed to determine 'delayed source term' are then described, including the various stages of phenomena and processes: fission product (FP) release from fuel , evaporation release from the surface of the pool, iodine mass transfer from fission gas bubble, FP deposition , and aerosol release from core-concrete interaction. The research programs to investigate the release and transport of FPs and aerosols in the reactor containment (i.e., in-containment source term) are not described in this report

  7. The PEC reactor. Safety analysis: Detailed reports

    Energy Technology Data Exchange (ETDEWEB)

    1988-01-01

    In the safety-analysis of the PEC Brasimone reactor (Italy), attention was focused on the role of plant-incident analysis during the design stage and the conclusions reached. The analysis regarded the following: thermohydraulic incidents at full power; incidents with the reactor shut down; reactivity incidents; core local faults; analysis of fuel-handling incidents; engineered safeguards and passive safety features; coolant leakage and sodium fires; research and development studies on the seismic behaviour of the PEC fast reactor; generalized sodium fire; severe accidents, accident sequences with shudown; reference accident. Both the theoretical and experimental analyses demonstrated the adequacy of the design of the PEC fast reactor, aimed at minimizing the consequences of a hypothetical disruptive core accident with mechanical energy release. It was shown that the containment barriers were sized correctly and that the residual heat from a disassembled core would be removed. The re-evaluation of the source term emphasized the conservative nature of the hypotheses assumed in the preliminary safety analysis for calculating the risk to the public.

  8. 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

  9. Study on CO{sub 2} Recovery System Design in Supercritical CO{sub 2} Cycle for SFR Application

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Min Seok; Jung, Hwa-Young; Lee, Jeong Ik [KAIST, Daejeon (Korea, Republic of)

    2016-10-15

    As a part of Sodium-cooled Fast Reactor (SFR) development in Korea, the supercritical CO{sub 2} (S-CO{sub 2}) Brayton cycle is considered as an alternative power conversion system to eliminate sodium-water reaction (SWR) when the current conventional steam Rankine cycle is utilized with SFR. The parasitic loss caused by the leakage flow should be minimized since this greatly influences the cycle efficiency. Thus, a simple model for estimating the critical flow in a turbo-machinery seal was developed to predict the leakage flow rate and calculate the required total mass of working fluid in a S-CO{sub 2} power system to minimize the parasitic loss. In this work, study on CO{sub 2} recovery system design was conducted by finding the suitable recovery point with the developed simple CO{sub 2} critical flow model and sensitivity analysis was performed on the power system performance with respect to multiple CO{sub 2} recovery process options. The study of a CO{sub 2} recovery system design was conducted to minimize the thermal efficiency losses caused by CO{sub 2} inventory recovery system. For the first step, the configuration of a seal was selected. A labyrinth seal has suitable features for the S-CO{sub 2} power cycle application. Then, thermal efficiency losses with different CO{sub 2} leak rate and recovery point were evaluated. To calculate the leak rate in turbo-machinery by using the developed CO{sub 2} critical flow model, the conditions of storage tank is set to be closer to the recovery point. After modifying the critical flow model appropriately, total mass flow rate of leakage flow was calculated. Finally, the CO{sub 2} recovery system design work was performed to minimize the loss of thermal efficiency. The suggested system is not only simple and intuitive but also has relatively very low additional work loss from the compressor than other considered systems. When each leak rate is set to the conventional leakage rate of 1 kg/s per seal, the minimum and

  10. Site investigation SFR. Fracture mineralogy and geochemistry of borehole sections sampled for groundwater chemistry and Eh. Results from boreholes KFR01, KFR08, KFR10, KFR19, KFR7A and KFR105

    Energy Technology Data Exchange (ETDEWEB)

    Sandstroem, Bjoern (WSP Sverige AB (Sweden)); Tullborg, Eva-Lena (Terralogica AB, Grabo (Sweden))

    2011-01-15

    This report is part of the complementary site investigations for the future expansion of SFR. The report presents the results obtained during a detailed mineralogical and geochemical study of fracture minerals in drill cores from borehole section sampled for groundwater chemistry and where downhole Eh measurements have been performed. The groundwater redox system comprises not only the water, but also the bedrock/fracture mineral system in contact with this water. It is thus important to gain knowledge of the solid phases in contact with the groundwater, i.e. the fracture minerals. The samples studied for mineralogy and geochemistry, here reported, were selected to represent the fracture surfaces in contact with the groundwater in the sampled borehole sections and will give input to the hydrogeochemical model (SFR SDM). The mineralogy was determined using SEM-EDS and XRD and the geochemistry of fracture filling material was analysed by ICP-AES and ICP-QMS. The most common fracture minerals in the samples are mixed layer clay (smectite-illite), illite, chlorite, calcite, quartz, adularia and albite. Other minerals identified in the borehole sections include laumontite, pyrite, barite, chalcopyrite, hematite, Fe-oxyhydroxide, muscovite, REE-carbonate, allanite, biotite, asphaltite, galena, sphalerite, arsenopyrite, uranium phosphate, uranium silicate, Y-Ca silicate, monazite, xenotime, harmotome and fluorite. There are no major differences between the fracture mineralogy of the investigated borehole sections from SFR and the fracture mineralogy of the Forsmark site investigation area. The four fracture mineral generations distinguished within the Forsmark site investigation are also found at SFR. However, some differences have been observed: 1) Barite and uranium minerals are more common in the SFR fractures, 2) clay minerals like mixed layer illite-smectite and illite dominates in contrast to Forsmark where corrensite is by far the most common clay mineral and, 3

  11. Site investigation SFR. Fracture mineralogy and geochemistry of borehole sections sampled for groundwater chemistry and Eh. Results from boreholes KFR01, KFR08, KFR10, KFR19, KFR7A and KFR105

    International Nuclear Information System (INIS)

    Sandstroem, Bjoern; Tullborg, Eva-Lena

    2011-01-01

    This report is part of the complementary site investigations for the future expansion of SFR. The report presents the results obtained during a detailed mineralogical and geochemical study of fracture minerals in drill cores from borehole section sampled for groundwater chemistry and where downhole Eh measurements have been performed. The groundwater redox system comprises not only the water, but also the bedrock/fracture mineral system in contact with this water. It is thus important to gain knowledge of the solid phases in contact with the groundwater, i.e. the fracture minerals. The samples studied for mineralogy and geochemistry, here reported, were selected to represent the fracture surfaces in contact with the groundwater in the sampled borehole sections and will give input to the hydrogeochemical model (SFR SDM). The mineralogy was determined using SEM-EDS and XRD and the geochemistry of fracture filling material was analysed by ICP-AES and ICP-QMS. The most common fracture minerals in the samples are mixed layer clay (smectite-illite), illite, chlorite, calcite, quartz, adularia and albite. Other minerals identified in the borehole sections include laumontite, pyrite, barite, chalcopyrite, hematite, Fe-oxyhydroxide, muscovite, REE-carbonate, allanite, biotite, asphaltite, galena, sphalerite, arsenopyrite, uranium phosphate, uranium silicate, Y-Ca silicate, monazite, xenotime, harmotome and fluorite. There are no major differences between the fracture mineralogy of the investigated borehole sections from SFR and the fracture mineralogy of the Forsmark site investigation area. The four fracture mineral generations distinguished within the Forsmark site investigation are also found at SFR. However, some differences have been observed: 1) Barite and uranium minerals are more common in the SFR fractures, 2) clay minerals like mixed layer illite-smectite and illite dominates in contrast to Forsmark where corrensite is by far the most common clay mineral and, 3

  12. JSFR design progress related to development of safety design criteria for Generation IV sodium-cooled fast reactors. (1) Overview

    International Nuclear Information System (INIS)

    Kamide, Hideki; Ando, Masato; Ito, Takaya

    2015-01-01

    JAEA, JAPC and MFBR have been conducting design study for the Japan Sodium-cooled Fast Reactor (JSFR), which is a design concept aiming at future commercial use as sustainable electric power source. As the result of the design study and R and D activity related the innovative technologies incorporated in the design in the Fast Reactor Cycle Technology Development (FaCT) project up to 2010, basic design concept of JSFR was established and its development process to the commercialization including construction and operation of a demonstration version of JSFR was outlined. JSFR is a looptype next generation sodium-cooled fast reactor (SFR), which is aiming at achieving development targets of Generation IV reactors concerning sustainability, safety and reliability, economics and proliferation resistance and physical protection by introducing the innovative technologies such as shortened high-chromium steel piping. The output power is assumed for the design study as 1,500 MWe for the commercial version and 750 MWe for the demonstration version. In FaCT phase I up to 2010, in order to evaluate feasibility to achieve the development targets, the design study has been conducted on the main components and systems. Since 2011, in order to contribute to the development of safety design criteria (SDC) and safety design guideline (SDG), which include the lessons learned from the TEPCO's Fukushima Dai-ichi nuclear power plants accident, in the frame work of Generation IV International Forum (GIF), the design study is focusing on the design measures against severe external events such as earthquake and tsunami. At the same time, the design study is going into detail and paying much attention to the maintenance and repair to make surer its feasibility. This paper summarizes the design concept of the demonstration version of JSFR in which progress of design work was incorporated for the safety issues on SDC and SDG of a SFR. (author)

  13. Cost benefit analysis of reactor safety systems

    International Nuclear Information System (INIS)

    Maurer, H.A.

    1984-01-01

    Cost/benefit analysis of reactor safety systems is a possibility appropriate to deal with reactor safety. The Commission of the European Communities supported a study on the cost-benefit or cost effectiveness of safety systems installed in modern PWR nuclear power plants. The following systems and their cooperation in emergency cases were in particular investigated in this study: the containment system (double containment), the leakage exhaust and control system, the annulus release exhaust system and the containment spray system. The benefit of a safety system is defined according to its contribution to the reduction of the radiological consequences for the environment after a LOCA. The analysis is so far performed in two different steps: the emergency core cooling system is considered to function properly, failure of the emergency core cooling system is assumed (with the possible consequence of core melt-down) and the results may demonstrate the evidence that striving for cost-effectiveness can produce a safer end result than the philosophy of safety at any cost. (orig.)

  14. Safety systems and safety analysis of the Qinshan phase III CANDU nuclear power plant

    International Nuclear Information System (INIS)

    Cai Jianping; Shen Sen; Barkman, N.

    1999-01-01

    The author introduces the Canadian nuclear reactor safety philosophy and the Qinshan Phase III CANDU NPP safety systems and safety analysis, which are designed and performed according to this philosophy. The concept of 'defence-in-depth' is a key element of the Canadian nuclear reactor safety philosophy. The design concepts of redundancy, diversity, separation, equipment qualification, quality assurance, and use of appropriate design codes and standards are adopted in the design. Four special safety systems as well as a set of reliable safety support systems are incorporated in the design of Qinshan phase III CANDU for accident mitigation. The assessment results for safety systems performance show that the fundamental safety criteria for public dose, and integrity of fuel, channels and the reactor building, are satisfied

  15. Project Safe. Gas related processes in SFR

    International Nuclear Information System (INIS)

    Moreno, L.

    2001-06-01

    The radionuclide release from the SFR repository caused by gas generation was calculated for different scenarios for three repository parts (Silo, BMA and 1BTF). The calculation cases are based on the way the gas escapes from the concrete structures. In the basic cases the gas escapes through the evacuation pipes in the concrete lid of the Silo, through existing gaps between the concrete walls and the lid in BMA, and through the concrete backfill surrounding the waste packages in 1BTF. These cases correspond to the situation that we expect to occur. Another category of cases corresponds to the situation where an initial fracture exists in the concrete structures. The fracture is assumed to exist at the bottom of the respective concrete structure in the Silo and BMA. For 1BTF the initial defect is represented by a fracture transversely crossing the section containing the steel drums with ashes. Other cases were also calculated with the purpose of studying some special situations. For example, the consequences of a silo repository without evacuation pipes and backfill in the interior of BMA. The radionuclide release, for some radionuclides, may be increased by several orders of magnitude when contaminated water is expelled by gas from the interior of the concrete structures. However, the impact on the total doses during the first thousands years after closure of the repository is limited. The total dose is dominated by the release of organic 14 C. Since the radionuclides are released to the coastal area during the first thousand years the dilution is considerable, which results in a very low dose

  16. Project Safe. Gas related processes in SFR

    Energy Technology Data Exchange (ETDEWEB)

    Moreno, L. [Royal Inst. of Technology, Stockholm (Sweden). Dept. of Chemical Engineering; Skagius, K.; Soedergren, S.; Wiborgh, M. [Kemakta Konsult AB, Stockholm (Sweden)

    2001-06-01

    The radionuclide release from the SFR repository caused by gas generation was calculated for different scenarios for three repository parts (Silo, BMA and 1BTF). The calculation cases are based on the way the gas escapes from the concrete structures. In the basic cases the gas escapes through the evacuation pipes in the concrete lid of the Silo, through existing gaps between the concrete walls and the lid in BMA, and through the concrete backfill surrounding the waste packages in 1BTF. These cases correspond to the situation that we expect to occur. Another category of cases corresponds to the situation where an initial fracture exists in the concrete structures. The fracture is assumed to exist at the bottom of the respective concrete structure in the Silo and BMA. For 1BTF the initial defect is represented by a fracture transversely crossing the section containing the steel drums with ashes. Other cases were also calculated with the purpose of studying some special situations. For example, the consequences of a silo repository without evacuation pipes and backfill in the interior of BMA. The radionuclide release, for some radionuclides, may be increased by several orders of magnitude when contaminated water is expelled by gas from the interior of the concrete structures. However, the impact on the total doses during the first thousands years after closure of the repository is limited. The total dose is dominated by the release of organic {sup 14}C. Since the radionuclides are released to the coastal area during the first thousand years the dilution is considerable, which results in a very low dose.

  17. Challenges on innovations of newly-developed safety analysis codes

    International Nuclear Information System (INIS)

    Yang, Yanhua; Zhang, Hao

    2016-01-01

    With the development of safety analysis method, the safety analysis codes meet more challenges. Three challenges are presented in this paper, which are mathematic model, code design and user interface. Combined with the self-reliance safety analysis code named COSINE, the ways of meeting these requirements are suggested, that is to develop multi-phases, multi-fields and multi-dimension models, to adopt object-oriented code design ideal and to improve the way of modeling, calculation control and data post-processing in the user interface.

  18. Challenges on innovations of newly-developed safety analysis codes

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Yanhua [Shanghai Jiao Tong Univ. (China). School of Nuclear Science and Engineering; Zhang, Hao [State Nuclear Power Software Development Center, Beijing (China). Beijing Future Science and Technology City

    2016-05-15

    With the development of safety analysis method, the safety analysis codes meet more challenges. Three challenges are presented in this paper, which are mathematic model, code design and user interface. Combined with the self-reliance safety analysis code named COSINE, the ways of meeting these requirements are suggested, that is to develop multi-phases, multi-fields and multi-dimension models, to adopt object-oriented code design ideal and to improve the way of modeling, calculation control and data post-processing in the user interface.

  19. Safety Analysis Report for Ignalina NPP

    International Nuclear Information System (INIS)

    Negrivoda, G.

    1997-01-01

    In December 1994 an agreement was signed between the European Bank for Reconstruction and Development and the Republic of Lithuania for the grant of 32.86 MECU for the safety Improvement at Ignalina NPP. One of the conditions for the provision of the grant, was a requirement for an in-depth analysis of the safety level at Ignalina NPP in the scope and according to the standards acceptable for a western nuclear power plant, and to publish a Safety Analysis Report (SAR). The report should investigate and analyze any factor that could limit a safe operation of the plant, and provide recommendations for actual safety improvements. According to the agreement, Lithuania had to finalize the SAR until 31 December, 1995. The bank has also organized and financed investigation of safety at Ignalina NPP and preparation of the SAR. EBRD made an agreement with Sweden's Vattenfall, which subcontracted well-known companies from Canada, USA, Germany, etc., and also the Russian Research and Development Institute of Power Engineering (NIKIET), reactor designer of Ignalina NPP. The SAR is a very comprehensive document and contains about 8000 pages of text, diagrams and tables. The main findings of the SAR are provided in the article. A large number of discrepancies with modern rules and western practices was detected, but they were not proved to be serious enough to require reactors shutdown. Based on the recommendations of the SAR Ignalina NPP has worked out Safety Improvement Program No. 2 (SIP-2), which is planned for three years and will cost 486 MLT. (author)

  20. Safety culture and accident analysis-A socio-management approach based on organizational safety social capital

    Energy Technology Data Exchange (ETDEWEB)

    Rao, Suman [Risk Analyst (India)]. E-mail: sumanashokrao@yahoo.co.in

    2007-04-11

    One of the biggest challenges for organizations in today's competitive business environment is to create and preserve a self-sustaining safety culture. Typically, Key drivers of safety culture in many organizations are regulation, audits, safety training, various types of employee exhortations to comply with safety norms, etc. However, less evident factors like networking relationships and social trust amongst employees, as also extended networking relationships and social trust of organizations with external stakeholders like government, suppliers, regulators, etc., which constitute the safety social capital in the Organization-seem to also influence the sustenance of organizational safety culture. Can erosion in safety social capital cause deterioration in safety culture and contribute to accidents? If so, how does it contribute? As existing accident analysis models do not provide answers to these questions, CAMSoC (Curtailing Accidents by Managing Social Capital), an accident analysis model, is proposed. As an illustration, five accidents: Bhopal (India), Hyatt Regency (USA), Tenerife (Canary Islands), Westray (Canada) and Exxon Valdez (USA) have been analyzed using CAMSoC. This limited cross-industry analysis provides two key socio-management insights: the biggest source of motivation that causes deviant behavior leading to accidents is 'Faulty Value Systems'. The second biggest source is 'Enforceable Trust'. From a management control perspective, deterioration in safety culture and resultant accidents is more due to the 'action controls' rather than explicit 'cultural controls'. Future research directions to enhance the model's utility through layering are addressed briefly.

  1. Construction safety and waste management an economic analysis

    CERN Document Server

    Li, Rita Yi Man

    2015-01-01

    This monograph presents an analysis of construction safety problems and on-site safety measures from an economist’s point of view. The book includes examples from both emerging countries, e.g. China and India, and developed countries, e.g. Australia and Hong Kong. Moreover, the author covers an analysis on construction safety knowledge sharing by means of updatable mobile technology such as apps in Androids and iOS platform mobile devices. The target audience comprises primarily researchers and experts in the field but the book may also be beneficial for graduate students.

  2. Using Addenda in Documented Safety Analysis Reports

    International Nuclear Information System (INIS)

    Swanson, D.S.; Thieme, M.A.

    2003-01-01

    This paper discusses the use of addenda to the Radioactive Waste Management Complex (RWMC) Documented Safety Analysis (DSA) located at the Idaho National Engineering and Environmental Laboratory (INEEL). Addenda were prepared for several systems and processes at the facility that lacked adequate descriptive information and hazard analysis in the DSA. They were also prepared for several new activities involving unreviewed safety questions (USQs). Ten addenda to the RWMC DSA have been prepared since the last annual update

  3. System safety analysis of an autonomous mobile robot

    International Nuclear Information System (INIS)

    Bartos, R.J.

    1994-01-01

    Analysis of the safety of operating and maintaining the Stored Waste Autonomous Mobile Inspector (SWAMI) II in a hazardous environment at the Fernald Environmental Management Project (FEMP) was completed. The SWAMI II is a version of a commercial robot, the HelpMate trademark robot produced by the Transitions Research Corporation, which is being updated to incorporate the systems required for inspecting mixed toxic chemical and radioactive waste drums at the FEMP. It also has modified obstacle detection and collision avoidance subsystems. The robot will autonomously travel down the aisles in storage warehouses to record images of containers and collect other data which are transmitted to an inspector at a remote computer terminal. A previous study showed the SWAMI II has economic feasibility. The SWAMI II will more accurately locate radioactive contamination than human inspectors. This thesis includes a System Safety Hazard Analysis and a quantitative Fault Tree Analysis (FTA). The objectives of the analyses are to prevent potentially serious events and to derive a comprehensive set of safety requirements from which the safety of the SWAMI II and other autonomous mobile robots can be evaluated. The Computer-Aided Fault Tree Analysis (CAFTA copyright) software is utilized for the FTA. The FTA shows that more than 99% of the safety risk occurs during maintenance, and that when the derived safety requirements are implemented the rate of serious events is reduced to below one event per million operating hours. Training and procedures in SWAMI II operation and maintenance provide an added safety margin. This study will promote the safe use of the SWAMI II and other autonomous mobile robots in the emerging technology of mobile robotic inspection

  4. Preliminary Safety Analysis Report for the Tokamak Physics Experiment

    International Nuclear Information System (INIS)

    Motloch, C.G.; Bonney, R.F.; Levine, J.D.; Masson, L.S.; Commander, J.C.

    1995-04-01

    This Preliminary Safety Analysis Report (PSAR), includes an indication of the magnitude of facility hazards, complexity of facility operations, and the stage of the facility life-cycle. It presents the results of safety analyses, safety assurance programs, identified vulnerabilities, compensatory measures, and, in general, the rationale describing why the Tokamak Physics Experiment (TPX) can be safely operated. It discusses application of the graded approach to the TPX safety analysis, including the basis for using Department of Energy (DOE) Order 5480.23 and DOE-STD-3009-94 in the development of the PSAR

  5. Reliability analysis of diverse safety logic systems of fast breeder reactor

    International Nuclear Information System (INIS)

    Ravi Kumar, Bh.; Apte, P.R.; Srivani, L.; Ilango Sambasivan, S.; Swaminathan, P.

    2006-01-01

    Safety Logic for Fast Breeder Reactor (FBR) is designed to initiate safety action against Design Basis Events. Based on the outputs of various processing circuits, Safety logic system drives the control rods of the shutdown system. So, Safety Logic system is classified as safety critical system. Therefore, reliability analysis has to be performed. This paper discusses the Reliability analysis of Diverse Safety logic systems of FBRs. For this literature survey on safety critical systems, system reliability approach and standards to be followed like IEC-61508 are discussed in detail. For Programmable Logic device based systems, Hardware Description Languages (HDL) are used. So this paper also discusses the Verification and Validation for HDLs. Finally a case study for the Reliability analysis of Safety logic is discussed. (author)

  6. ALMA + VLT observations of a damped Lyman-α absorbing galaxy: massive, wide CO emission, gas-rich but with very low SFR

    Science.gov (United States)

    Møller, P.; Christensen, L.; Zwaan, M. A.; Kanekar, N.; Prochaska, J. X.; Rhodin, N. H. P.; Dessauges-Zavadsky, M.; Fynbo, J. P. U.; Neeleman, M.; Zafar, T.

    2018-03-01

    We are undertaking an Atacama Large Millimeter Array survey of molecular gas in galaxies selected for their strong H I absorption, so-called damped Lyα absorber (DLA)/sub-DLA galaxies. Here, we report CO(2-1) detection from a DLA galaxy at z = 0.716. We also present optical and near-infrared (NIR) spectra of the galaxy revealing [O II], Hα, and [N II] emission lines shifted by ˜170 km s-1 relative to the DLA, and providing an oxygen abundance 3.2 times solar, similar to the absorption metallicity. We report low unobscured SFR˜1 M⊙ yr-1 given the large reservoir of molecular gas, and also modest obscured SFR =4.5_{-2.6}^{+4.4} M⊙ yr-1 based on far-IR and sub-millimetre data. We determine mass components of the galaxy: log[M*/M_{&sun} ]= 10.80^{+0.07}_{-0.14}, log[Mmol-gas/M⊙] = 10.37 ± 0.04, and log[Mdust/M_{⊙} ]= 8.45^{+0.10}_{-0.30}. Surprisingly, this H I absorption-selected galaxy has no equivalent objects in CO surveys of flux-selected samples. The galaxy falls off current scaling relations for the star formation rate (SFR) to molecular gas mass and CO Tully-Fisher relation. Detailed comparison of kinematical components of the absorbing, ionized, and molecular gas, combined with their spatial distribution, suggests that part of the CO gas is both kinematically and spatially decoupled from the main galaxy. It is thus possible that a major starburst in the past could explain the wide CO profile as well as the low SFR. Support for this also comes from the spectral energy distribution favouring an instantaneous burst of age ≈0.5 Gyr. Our survey will establish whether flux-selected surveys of molecular gas are missing a key stage in the evolution of galaxies and their conversion of gas to stars.

  7. Uncertainty analysis for Ulysses safety evaluation report

    International Nuclear Information System (INIS)

    Frank, M.V.

    1991-01-01

    As part of the effort to review the Ulysses Final Safety Analysis Report and to understand the risk of plutonium release from the Ulysses spacecraft General Purpose Heat Source---Radioisotope Thermal Generator (GPHS-RTG), the Interagency Nuclear Safety Review Panel (INSRP) and the author performed an integrated, quantitative analysis of the uncertainties of the calculated risk of plutonium release from Ulysses. Using state-of-art probabilistic risk assessment technology, the uncertainty analysis accounted for both variability and uncertainty of the key parameters of the risk analysis. The results show that INSRP had high confidence that risk of fatal cancers from potential plutonium release associated with calculated launch and deployment accident scenarios is low

  8. COLD-SAT feasibility study safety analysis

    Science.gov (United States)

    Mchenry, Steven T.; Yost, James M.

    1991-01-01

    The Cryogenic On-orbit Liquid Depot-Storage, Acquisition, and Transfer (COLD-SAT) satellite presents some unique safety issues. The feasibility study conducted at NASA-Lewis desired a systems safety program that would be involved from the initial design in order to eliminate and/or control the inherent hazards. Because of this, a hazards analysis method was needed that: (1) identified issues that needed to be addressed for a feasibility assessment; and (2) identified all potential hazards that would need to be controlled and/or eliminated during the detailed design phases. The developed analysis method is presented as well as the results generated for the COLD-SAT system.

  9. Guidance for preparation of safety analysis reports for nonreactor facilities and operations

    International Nuclear Information System (INIS)

    1992-01-01

    Department of Energy (DOE) Orders 5480.23, ''Nuclear Safety Analysis Reports,'' and 5481.1B, ''Safety Analysis and Review System'' require the preparation of appropriate safety analyses for each DOE operation and subsequent significant modifications including decommissioning, and independent review of each safety analysis. The purpose of this guide is to assist in the preparation and review of safety documentation for Oak Ridge Field Office (OR) nonreactor facilities and operation. Appendix A lists DOE Orders, NRC Regulatory Guides and other documents applicable to the preparation of safety analysis reports

  10. Overview of major HZDR developments for fast reactor analysis

    Energy Technology Data Exchange (ETDEWEB)

    Merk, Bruno, E-mail: b.merk@hzdr.de; Glivici-Cotruţă, V.; Duerigen, S.; Rohde, U.; Kliem, S.

    2013-12-15

    The upgrading of the DYN3D code for the application for fast reactors is described. After extension and validation, a diverse code with the possibility for steady state and transient core analysis on the basis of coupled thermal hydraulics/neutronics calculations is available. The work on the use of fine distributed moderating material in SFR cores is discussed with the target on enhancing the feedback effects in SFR cores without influencing the operational. Newly developed analytical solutions without separation of space and time for the analysis of ADS experiments are shown with good agreement for the YALINA experiment. The analytical solutions are a very promising tool for the development of a new method for the analysis of ADS experiments.

  11. Safety analysis and risk assessment handbook

    International Nuclear Information System (INIS)

    Peterson, V.L.; Colwell, R.G.; Dickey, R.L.

    1997-01-01

    This Safety Analysis and Risk Assessment Handbook (SARAH) provides guidance to the safety analyst at the Rocky Flats Environmental Technology Site (RFETS) in the preparation of safety analyses and risk assessments. Although the older guidance (the Rocky Flats Risk Assessment Guide) continues to be used for updating the Final Safety Analysis Reports developed in the mid-1980s, this new guidance is used with all new authorization basis documents. With the mission change at RFETS came the need to establish new authorization basis documents for its facilities, whose functions had changed. The methodology and databases for performing the evaluations that support the new authorization basis documents had to be standardized, to avoid the use of different approaches and/or databases for similar accidents in different facilities. This handbook presents this new standardized approach. The handbook begins with a discussion of the requirements of the different types of authorization basis documents and how to choose the one appropriate for the facility to be evaluated. It then walks the analyst through the process of identifying all the potential hazards in the facility, classifying them, and choosing the ones that need to be analyzed further. It then discusses the methods for evaluating accident initiation and progression and covers the basic steps in a safety analysis, including consequence and frequency binning and risk ranking. The handbook lays out standardized approaches for determining the source terms of the various accidents (including airborne release fractions, leakpath factors, etc.), the atmospheric dispersion factors appropriate for Rocky Flats, and the methods for radiological and chemical consequence assessments. The radiological assessments use a radiological open-quotes templateclose quotes, a spreadsheet that incorporates the standard values of parameters, whereas the chemical assessments use the standard codes ARCHIE and ALOHA

  12. Design characteristics of pantograph type in vessel fuel handling system in SFR

    International Nuclear Information System (INIS)

    Kim, S. H.; Koo, G. H.

    2012-01-01

    The pantograph type in vessel fuel handling system in a sodium cooled fast reactor (SFR), which requires installation space for the slot in the upper internal structure attached under the rotating plug, is composed of an in vessel transfer machine (IVTM), a single rotating plug, in vessel storage, and a fuel transfer port (FTP). The pantograph type IVTM can exchange fuel assemblies through a slot, the design requirement of which should be essentially considered in the design of the in vessel fuel handling system. In addition, the spent fuel assemblies temporarily stored in the in vessel storage of the reactor vessel are removed to the outside of the reactor vessel through the FTP. The fuel transfer basket is then provided in the FTP, and a fuel transfer is performed by using it. In this study, the design characteristics for a pantograph type in vessel fuel handling system are reviewed, and the preconceptual designs are studied

  13. Design characteristics of pantograph type in vessel fuel handling system in SFR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S. H.; Koo, G. H. [KAERI, Daejeon (Korea, Republic of)

    2012-10-15

    The pantograph type in vessel fuel handling system in a sodium cooled fast reactor (SFR), which requires installation space for the slot in the upper internal structure attached under the rotating plug, is composed of an in vessel transfer machine (IVTM), a single rotating plug, in vessel storage, and a fuel transfer port (FTP). The pantograph type IVTM can exchange fuel assemblies through a slot, the design requirement of which should be essentially considered in the design of the in vessel fuel handling system. In addition, the spent fuel assemblies temporarily stored in the in vessel storage of the reactor vessel are removed to the outside of the reactor vessel through the FTP. The fuel transfer basket is then provided in the FTP, and a fuel transfer is performed by using it. In this study, the design characteristics for a pantograph type in vessel fuel handling system are reviewed, and the preconceptual designs are studied.

  14. Multivariate time series analysis of SafetyNet data. SafetyNet, Building the European Road Safety Observatory, Workpackage 7, Deliverable 7.7.

    NARCIS (Netherlands)

    Commandeur, J.J.F. Bijleveld, F.D. & Bergel, R.

    2009-01-01

    This deliverable provides an application of theories and methods documented in Deliverables 7.4 and 7.5 of work package 7 of the SafetyNet project. In this deliverable, use of select analysis techniques is demonstrated through real world road safety analysis problems, using aggregate data which may

  15. Engineered safeguards and passive safety features (safety analysis detailed report no. 6)

    Energy Technology Data Exchange (ETDEWEB)

    1988-01-15

    The Safety-Analysis Summary lists the reactor's safety aspects for passive and active prevention of severe accidents and mitigation of accident consequences, i.e., intrinsic and passive protections of the plant; intrinsic and passive protections of the core; inherent decay-heat removal systems; rapid-shutdown systems; four physical containment barriers. This report goes into further details regarding some of this aspects.

  16. An investigation of sodium–CO{sub 2} interaction byproduct cleaning agent for SFR coupled with S-CO{sub 2} Brayton cycle

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Hwa-Young, E-mail: jhy0523@kaist.ac.kr [Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 373-1 Guseong-dong Yuseong-gu, Daejeon 305-701 (Korea, Republic of); Division of SFR NSSS System Design, Korea Atomic Energy Research Institute, 111, Daedeok-daero 989beon-gil, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Lee, Jeong Ik, E-mail: jeongiklee@kaist.ac.kr [Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 373-1 Guseong-dong Yuseong-gu, Daejeon 305-701 (Korea, Republic of); Wi, Myung-Hwan, E-mail: mhwi@kaeri.re.kr [Division of SFR NSSS System Design, Korea Atomic Energy Research Institute, 111, Daedeok-daero 989beon-gil, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Ahn, Hong Joo, E-mail: ahjoo@kaeri.re.kr [Division of Nuclear Chemistry Research, Korea Atomic Energy Research Institute, 111, Daedeok-daero 989beon-gil, Yuseong-gu, Daejeon 305-353 (Korea, Republic of)

    2016-02-15

    Highlights: • Study on cleaning agent was conducted to remove Na–CO{sub 2} interaction byproducts. • Screening criteria to select candidate substances as cleaning agents were suggested. • The mixtures of Na{sub 2}CO{sub 3} with NaBrO{sub 3}, NaClO{sub 3}, or NaBF{sub 4} were thermally analyzed with the TG/DTA studies. • Three candidate substances decomposed before 600 °C and did not react with Na{sub 2}CO{sub 3}. - Abstract: One of the promising future nuclear energy systems, the Sodium-cooled Fast Reactor (SFR) has been actively developed internationally. Recently, to improve safety and economics of a SFR further, coupling supercritical CO{sub 2} power cycle was suggested. However, there can be a chemical reaction between sodium and CO{sub 2} at high temperature (more than 400 °C) when the pressure boundary fails in a sodium–CO{sub 2} heat exchanger. To ensure the performance of such a system, it is important to employ a cleaning agent to recover the system back to normal condition after the reaction. When sodium and CO{sub 2} react, solid and gaseous reaction products such as sodium carbonate (Na{sub 2}CO{sub 3}) and carbon monoxide (CO) appear. Since most of solid reaction products are hard and can deteriorate system performance, quick removal of solid reaction products becomes very important for economic performance of the system. Thus, the authors propose the conceptual method to remove the byproducts with a chemical reaction at high temperature. The chemical reaction will take place between the reaction byproducts and a cleaning agent while the cleaning agent is inert with sodium. Thus, various sodium-based compounds were first investigated and three candidate substances satisfying several criteria were selected; sodium bromate (NaBrO{sub 3}), sodium chlorate (NaClO{sub 3}), and sodium tetrafluoroborate (NaBF{sub 4}). The selected substances were thermally analyzed with the TG/DTA studies. Unfortunately, it was revealed that all candidate

  17. Computer aided safety analysis

    International Nuclear Information System (INIS)

    1988-05-01

    The document reproduces 20 selected papers from the 38 papers presented at the Technical Committee/Workshop on Computer Aided Safety Analysis organized by the IAEA in co-operation with the Institute of Atomic Energy in Otwock-Swierk, Poland on 25-29 May 1987. A separate abstract was prepared for each of these 20 technical papers. Refs, figs and tabs

  18. Holistic safety analysis for advanced nuclear power plants

    International Nuclear Information System (INIS)

    Alvarenga, M.A.B.; Guimaraes, A.C.F.

    1992-01-01

    This paper reviews the basic methodology of safety analysis used in the ANGRA-I and ANGRA-II nuclear power plants, its weaknesses, the problems with public acceptance of the risks, the future of the nuclear energy in Brazil, as well as recommends a new methodology, HOLISTIC SAFETY ANALYSIS, to be used both in the design and licensing phases, for advanced reactors. (author)

  19. Proposed Advanced Reactor Adaptation of the Standard Review Plan NUREG-0800 Chapter 4 (Reactor) for Sodium-Cooled Fast Reactors and Modular High-Temperature Gas-Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Belles, Randy [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Poore, III, Willis P. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Flanagan, George F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Holbrook, Mark [Idaho National Lab. (INL), Idaho Falls, ID (United States); Moe, Wayne [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sofu, Tanju [Argonne National Lab. (ANL), Argonne, IL (United States)

    2017-03-01

    This report proposes adaptation of the previous regulatory gap analysis in Chapter 4 (Reactor) of NUREG 0800, Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light Water Reactor] Edition. The proposed adaptation would result in a Chapter 4 review plan applicable to certain advanced reactors. This report addresses two technologies: the sodium-cooled fast reactor (SFR) and the modular high temperature gas-cooled reactor (mHTGR). SRP Chapter 4, which addresses reactor components, was selected for adaptation because of the possible significant differences in advanced non-light water reactor (non-LWR) technologies compared with the current LWR-based description in Chapter 4. SFR and mHTGR technologies were chosen for this gap analysis because of their diverse designs and the availability of significant historical design detail.

  20. Special characteristics of the safety analysis of HWRs

    International Nuclear Information System (INIS)

    Kugler, G.

    1980-01-01

    Two lectures are presented in this report. The CANDU-PHW reactor is used as a model for discussion. The first lecture describes the distinctive features of the CANDU reactor, and how they impact on reactor safety. In the second lecture the Canadian safety philosophy, the safety design objective, and other selected topics on reactor safety analysis are discussed. The material in this report was selected with a view to assisting those not familiar with the CANDU heavy water reactor design in evaluating the distinctive safety aspects of these reactors. (orig./RW)

  1. SCALE 5: Powerful new criticality safety analysis tools

    International Nuclear Information System (INIS)

    Bowman, Stephen M.; Hollenbach, Daniel F.; Dehart, Mark D.; Rearden, Bradley T.; Gauld, Ian C.; Goluoglu, Sedat

    2003-01-01

    Version 5 of the SCALE computer software system developed at Oak Ridge National Laboratory, scheduled for release in December 2003, contains several significant new modules and sequences for criticality safety analysis and marks the most important update to SCALE in more than a decade. This paper highlights the capabilities of these new modules and sequences, including continuous energy flux spectra for processing multigroup problem-dependent cross sections; one- and three-dimensional sensitivity and uncertainty analyses for criticality safety evaluations; two-dimensional flexible mesh discrete ordinates code; automated burnup-credit analysis sequence; and one-dimensional material distribution optimization for criticality safety. (author)

  2. Support analysis for safety analysis development for CANDU nuclear power plant

    International Nuclear Information System (INIS)

    Bedreaga, L.; Florescu, Gh.; Apostol, M.; Nitoi, M.

    2004-01-01

    Probabilistic Safety Assessment analysis (PSA) is a technique used to assess the safety of a nuclear power plant. Assessments of the nuclear plant systems/components from safety point of view consist in accomplishment of a lot of support analyses that are the base for the main analysis, in order to evaluate the impact of occurrences of abnormal states for these systems. Evaluation of initiating events frequency and components failure rate is based on underlying probabilistic theory and mathematic statistics. Some of these analyses are detailed analyses and are known very well in PSA. There are also some analyses, named support analyses for PSA, which are very important but less applicable because they involve a huge human effort and hardware facilities to accomplish. The usual methods applicable in PSA such as input data extracted from the specific documentation (operation procedures, testing procedures, maintenance procedures and so on) or conservative evaluation provide a high level of uncertainty for both input and output data. The paper describes support analysis required to improve the certainty level in evaluation of reliability parameters and also in the final results (either risk, reliability or safety assessment). (author)

  3. Preparing a Safety Analysis Report using the building block approach

    International Nuclear Information System (INIS)

    Herrington, C.C.

    1990-01-01

    The credibility of the applicant in a licensing proceeding is severely impacted by the quality of the license application, particularly the Safety Analysis Report. To ensure the highest possible credibility, the building block approach was devised to support the development of a quality Safety Analysis Report. The approach incorporates a comprehensive planning scheme that logically ties together all levels of the investigation and provides the direction necessary to prepare a superior Safety Analysis Report

  4. Utilization of the MCNP-3A code for criticality safety analysis

    International Nuclear Information System (INIS)

    Maragni, M.G.; Moreira, J.M.L.

    1996-01-01

    In the last decade, Brazil started to operate facilities for processing and storing uranium in different forms. The necessity of criticality safety analysis appeared in the design phase of the uranium pilot process plants and also in the licensing of transportation and storage of fissile materials. The 2-MW research reactor and the Angra I power plant also required criticality safety assessments because their spent-fuel storage was approaching full-capacity utilization. The criticality safety analysis in Brazil has been based on KENO IV code calculations, which present some difficulties for correct geometry representation. The MCNP-3A code is not reported to be used frequently for criticality safety analysis in Brazil, but its good geometry representation makes it a possible tool for treating problems of complex geometry. A set of benchmark tests was performed to verify its applicability for criticality safety analysis in Brazil. This paper presents several benchmark tests aimed at selecting a set of options available in the MCNP-3A code that would be adequate for criticality safety analysis. The MCNP-3A code is also compared with the KENO-IV code regarding its performance for criticality safety analysis

  5. Safety analysis of tritium processing system based on PHA

    International Nuclear Information System (INIS)

    Fu Wanfa; Luo Deli; Tang Tao

    2012-01-01

    Safety analysis on primary confinement of tritium processing system for TBM was carried out with Preliminary Hazard Analysis. Firstly, the basic PHA process was given. Then the function and safe measures with multiple confinements about tritium system were described and analyzed briefly, dividing the two kinds of boundaries of tritium transferring through, that are multiple confinement systems division and fluid loops division. Analysis on tritium releasing is the key of PHA. Besides, PHA table about tritium releasing was put forward, the causes and harmful results being analyzed, and the safety measures were put forward also. On the basis of PHA, several kinds of typical accidents were supposed to be further analyzed. And 8 factors influencing the tritium safety were analyzed, laying the foundation of evaluating quantitatively the safety grade of various nuclear facilities. (authors)

  6. A Post Closure Safety Assessment for Radioactive Wastes from Advanced nuclear fuel Cycle

    International Nuclear Information System (INIS)

    Kang, Chul Hyung; Hwang, Yong Soo

    2010-01-01

    KAERI has developed the KIEP-21 (Korean, Innovative, Environmentally Friendly, and Proliferation Resistant System for the 21st Century). It is an advanced nuclear fuel cycle option with a pyro-process and a GEN-IV SFR. A pyro-process consists of two distinctive processes, an electrolytic reduction process and an electro-refining and winning process. When the pyro-process is applied, it generates five streams of wastes. To compare pyro-process advantage over the direct disposal of Spent Nuclear Fuel (SNF), the PWR SNF of the 45,000 MWD burn-up has been assumed. A safety assessment model for pyro-process wastes and representative results are presented in this report

  7. Advanced ADS concepts

    International Nuclear Information System (INIS)

    Slessarev, I.

    2001-01-01

    Current Nuclear Power has already proved its ability to be used as one of the principal means of the large scale energy production throughout the world. There is the widely spread opinion that many of the problems can be resolved if LWR's would be able to close its fuel cycle (MOX-strategy). Really, closure of fuel cycle (if it could be realised technically in full degree) leads to the reduction of the TRU-accumulation rate and of the most worrying long-lived fuel waste by factor of 10 -30 and, hence, to facilitation the waste repository problems. Analysis shows that one of the most important problems with MOX-type fuelled LWR exploitation is its potential safety degradation due to corresponding degradation of some principal physical parameters. Non-favourable feed back effects, delayed neutron fraction reduction, etc. lead, probably, to some important constraints in fuel multirecycling fraction. However, if the LWR's fuel cycle closure would be even realised, other, not less important, key-issues will rest unresolved or even more aggravated such as the economical competitiveness, weapons material proliferation, fuel resources, natural safety level, etc. Possible replacement of LWR's park by Sodium Fast Reactors (SFR) is able to change some accents in NP acceptance, however, can not change this situation drastically. Really, the potential of the waste long-term toxicity reduction is slightly favourable for SFR's than for LWRs (by the factor of 1.5 [1,2]) and this benefit is indebted mostly due to the higher fuel burnup potential of SFR's. Hence, the waste toxicity reduction factor, when SFR fuel cycle will be closed, is expecting to be in the interval 20-50 and, hence, it is slightly better than for LWRs. Nevertheless, it is important to mention that the closure of fast reactors fuel cycles does not lead to important degradation of safety physics. TRU equilibrium inventory is one of penalising factors of fast reactors: the total TRU inventory in NP, based on SFR

  8. A Study of Time Response for Safety-Related Operator Actions in Non-LOCA Safety Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Min Seok; Lee, Sang Seob; Park, Min Soo; Lee, Gyu Cheon; Kim, Shin Whan [KEPCO E and C Company, Daejeon (Korea, Republic of)

    2014-10-15

    The classification of initiating events for safety analysis report (SAR) chapter 15 is categorized into moderate frequency events (MF), infrequent events (IF), and limiting faults (LF) depending on the frequency of its occurrence. For the non-LOCA safety analysis with the purpose to get construction or operation license, however, it is assumed that the operator response action to mitigate the events starts at 30 minutes after the initiation of the transient regardless of the event categorization. Such an assumption of corresponding operator response time may have over conservatism with the MF and IF events and results in a decrease in the safety margin compared to its acceptance criteria. In this paper, the plant conditions (PC) are categorized with the definitions in SAR 15 and ANS 51.1. Then, the consequence of response for safety-related operator action time is determined based on the PC in ANSI 58.8. The operator response time for safety analysis regarding PC are reviewed and suggested. The clarifying alarm response procedure would be required for the guideline to reduce the operator response time when the alarms indicate the occurrence of the transient.

  9. Models for dose assessments. Models adapted to the SFR-area, Sweden

    Energy Technology Data Exchange (ETDEWEB)

    Karlsson, Sara; Bergstroem, U.; Meili, M. [Studsvik Eco and Safety AB, Nykoeping (Sweden)

    2001-10-01

    This report presents a model system created to be used to predict dose rates to the most exposed individuals in case of a long-term release of radionuclides from the Final repository for radioactive operational waste (SFR) in Forsmark, Sweden. The system accounts for an underground point source emitting radionuclides to the biosphere,their transport and distribution in various ecosystem types, their uptake by various biota, and calculation of doses to man from a multitude of exposure pathways. Long-term aspects include the consideration of processes at geological time scales, such as land uplift and conversion of marine sediments to arable land. Model parameters are whenever possible based on local conditions and recent literature. The system has been used for simulations based on geospheric releases varying over time of a mixture of radionuclides. Here, the models have been subjected to various test scenarios, including different radionuclide entry points and sorption properties. Radionuclides released from SFR are efficiently diluted to low concentrations in the water of the coastal model. A large fraction of the nuclides leave the Model Area quickly as a consequence of the rapid water turnover. The total amount of radionuclides in water is about the same independent of particle affinity (K{sub d} ), and at most some percents of the amounts retained in the sediments for some time. The latter is also true for the lake model when releases of radionuclides to the water is simulated. The role of sediments as a radionuclide source seems of minor importance in lakes at least for long-term radiation doses. Modelling the sediments as a source of radionuclides most of the 'low K{sub d} radionuclides' will leave the lake whereas the 'high K{sub d} nuclides' are still present within the deeper sediments after 1 000 years. The amount of 'low K{sub d} radionuclides' present in the water and on suspended matter are about 8x10{sup -5} of the

  10. Models for dose assessments. Models adapted to the SFR-area, Sweden

    International Nuclear Information System (INIS)

    Karlsson, Sara; Bergstroem, U.; Meili, M.

    2001-10-01

    This report presents a model system created to be used to predict dose rates to the most exposed individuals in case of a long-term release of radionuclides from the Final repository for radioactive operational waste (SFR) in Forsmark, Sweden. The system accounts for an underground point source emitting radionuclides to the biosphere,their transport and distribution in various ecosystem types, their uptake by various biota, and calculation of doses to man from a multitude of exposure pathways. Long-term aspects include the consideration of processes at geological time scales, such as land uplift and conversion of marine sediments to arable land. Model parameters are whenever possible based on local conditions and recent literature. The system has been used for simulations based on geospheric releases varying over time of a mixture of radionuclides. Here, the models have been subjected to various test scenarios, including different radionuclide entry points and sorption properties. Radionuclides released from SFR are efficiently diluted to low concentrations in the water of the coastal model. A large fraction of the nuclides leave the Model Area quickly as a consequence of the rapid water turnover. The total amount of radionuclides in water is about the same independent of particle affinity (K d ), and at most some percents of the amounts retained in the sediments for some time. The latter is also true for the lake model when releases of radionuclides to the water is simulated. The role of sediments as a radionuclide source seems of minor importance in lakes at least for long-term radiation doses. Modelling the sediments as a source of radionuclides most of the 'low K d radionuclides' will leave the lake whereas the 'high K d nuclides' are still present within the deeper sediments after 1 000 years. The amount of 'low K d radionuclides' present in the water and on suspended matter are about 8x10 -5 of the initial inventory in the sediments. For 'high K d nuclides

  11. Development of safety analysis technology for LMR

    International Nuclear Information System (INIS)

    Hahn, Do Hee; Kwon, Y. M.; Suk, S. D.

    2002-05-01

    In the present study, the KALIMER safety analysis has been made for the transients considered in the design concept, hypothetical core disruptive accident (HCDA), and containment performance with the establishment of the design basis. Such analyses have not been possible without the computer code improvement, and the experience attained during this research period must have greatly contributed to the achievement of the self reliance in the domestic technology establishment on the safety analysis areas of the conceptual design. The safety analysis codes have been improved to extend their applicable ranges for detailed conceptual design, and a basic computer code system has been established for HCDA analysis. A code-to-code comparison analysis has been performed as a part of code verification attempt, and the leading edge technology of JNC also has been brought for the technology upgrade. In addition, the research and development on the area of the database establishment has been made for the efficient and systematic project implementation of the conceptual design, through performances on the development of a project scheduling management, integration of the individually developed technology, establishment of the product database, and so on, taking into account coupling of the activities conducted in each specific area

  12. Preclosure Safety Analysis Guide

    International Nuclear Information System (INIS)

    D.D. Orvis

    2003-01-01

    A preclosure safety analysis (PSA) is a required element of the License Application (LA) for the high- level radioactive waste repository at Yucca Mountain. This guide provides analysts and other Yucca Mountain Repository Project (the Project) personnel with standardized methods for developing and documenting the PSA. The definition of the PSA is provided in 10 CFR 63.2, while more specific requirements for the PSA are provided in 10 CFR 63.112, as described in Sections 1.2 and 2. The PSA requirements described in 10 CFR Part 63 were developed as risk-informed performance-based regulations. These requirements must be met for the LA. The PSA addresses the safety of the Geologic Repository Operations Area (GROA) for the preclosure period (the time up to permanent closure) in accordance with the radiological performance objectives of 10 CFR 63.111. Performance objectives for the repository after permanent closure (described in 10 CFR 63.113) are not mentioned in the requirements for the PSA and they are not considered in this guide. The LA will be comprised of two phases: the LA for construction authorization (CA) and the LA amendment to receive and possess (R and P) high-level radioactive waste (HLW). PSA methods must support the safety analyses that will be based on the differing degrees of design detail in the two phases. The methods described herein combine elements of probabilistic risk assessment (PRA) and deterministic analyses that comprise a risk-informed performance-based safety analysis. This revision to the PSA guide was prepared for the following objectives: (1) To correct factual and typographical errors. (2) To provide additional material suggested from reviews by the Project, the U.S. Department of Energy (DOE), and U.S. Nuclear Regulatory Commission (NRC) Staffs. (3) To update material in accordance with approaches and/or strategies adopted by the Project. In addition, a principal objective for the planned revision was to ensure that the methods and

  13. Analysis of high burnup fuel safety issues

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chan Bock; Kim, D. H.; Bang, J. G.; Kim, Y. M.; Yang, Y. S.; Jung, Y. H.; Jeong, Y. H.; Nam, C.; Baik, J. H.; Song, K. W.; Kim, K. S

    2000-12-01

    Safety issues in steady state and transient behavior of high burnup LWR fuel above 50 - 60 MWD/kgU were analyzed. Effects of burnup extension upon fuel performance parameters was reviewed, and validity of both the fuel safety criteria and the performance analysis models which were based upon the lower burnup fuel test results was analyzed. It was found that further tests would be necessary in such areas as fuel failure and dispersion for RIA, and high temperature cladding corrosion and mechanical deformation for LOCA. Since domestic fuels have been irradiated in PWR up to burnup higher than 55 MWD/kgU-rod. avg., it can be said that Korea is in the same situation as the other countries in the high burnup fuel safety issues. Therefore, necessary research areas to be performed in Korea were derived. Considering that post-irradiation examination(PIE) for the domestic fuel of burnup higher than 30 MWD/kgU has not been done so far at all, it is primarily necessary to perform PIE for high burnup fuel, and then simulation tests for RIA and LOCA could be performed by using high burnup fuel specimens. For the areas which can not be performed in Korea, international cooperation will be helpful to obtain the test results. With those data base, safety of high burnup domestic fuels will be confirmed, current fuel safety criteria will be re-evaluated, and finally transient high burnup fuel behavior analysis technology will be developed through the fuel performance analysis code development.

  14. Analysis of high burnup fuel safety issues

    International Nuclear Information System (INIS)

    Lee, Chan Bock; Kim, D. H.; Bang, J. G.; Kim, Y. M.; Yang, Y. S.; Jung, Y. H.; Jeong, Y. H.; Nam, C.; Baik, J. H.; Song, K. W.; Kim, K. S

    2000-12-01

    Safety issues in steady state and transient behavior of high burnup LWR fuel above 50 - 60 MWD/kgU were analyzed. Effects of burnup extension upon fuel performance parameters was reviewed, and validity of both the fuel safety criteria and the performance analysis models which were based upon the lower burnup fuel test results was analyzed. It was found that further tests would be necessary in such areas as fuel failure and dispersion for RIA, and high temperature cladding corrosion and mechanical deformation for LOCA. Since domestic fuels have been irradiated in PWR up to burnup higher than 55 MWD/kgU-rod. avg., it can be said that Korea is in the same situation as the other countries in the high burnup fuel safety issues. Therefore, necessary research areas to be performed in Korea were derived. Considering that post-irradiation examination(PIE) for the domestic fuel of burnup higher than 30 MWD/kgU has not been done so far at all, it is primarily necessary to perform PIE for high burnup fuel, and then simulation tests for RIA and LOCA could be performed by using high burnup fuel specimens. For the areas which can not be performed in Korea, international cooperation will be helpful to obtain the test results. With those data base, safety of high burnup domestic fuels will be confirmed, current fuel safety criteria will be re-evaluated, and finally transient high burnup fuel behavior analysis technology will be developed through the fuel performance analysis code development

  15. Progress reports for Gen IV sodium fast reactor activities FY 2007

    International Nuclear Information System (INIS)

    Cahalan, J. E.; Tentner, A. M.

    2007-01-01

    An important goal of the US DOE Sodium Fast Reactor (SFR) program is to develop the technology necessary to increase safety margins in future fast reactor systems. Although no decision has been made yet about who will build the next demonstration fast reactor, it seems likely that the construction team will include a combination of international companies, and the safety design philosophy for the reactor will reflect a consensus of the participating countries. A significant amount of experience in the design and safety analysis of Sodium Fast Reactors (SFR) using oxide fuel has been developed in both Japan and France during last few decades. In the US, the traditional approach to reactor safety is based on the principle of defense-in-depth, which is usually expressed in physical terms as multiple barriers to release of radioactive material (e.g. cladding, reactor vessel, containment building), but it is understood that the 'barriers' may consist of active systems or even procedures. As implemented in a reactor design, defense-in-depth is classed in levels of safety. Level 1 includes measures to specify and build a reliable design with significant safety margins that will perform according to the intentions of the designers. Level 2 consists of additional design measures, usually active systems, to protect against unlikely accidental events that may occur during the life of the plant. Level 3 design measures are intended to protect the public in the event of an extremely unlikely accident not foreseen to occur during the plant's life. All of the design measures that make up the first three levels of safety are within the design basis of the plant. Beyond Level 3, and beyond the normal design basis, there are accidents that are not expected to occur in a whole generation of plants, and it is in this class that severe accidents, i.e. accidents involving core melting, are included. Beyond design basis measures to address severe accidents are usually identified as being

  16. LOCA analysis of SCWR-M with passive safety system

    Energy Technology Data Exchange (ETDEWEB)

    Liu, X.J., E-mail: xiaojingliu@sjtu.edu.cn [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, 800 Dong Chuan Road, Shanghai 200240 (China); Fu, S.W. [Navy University of Engineering, Wuhan, Hubei (China); Xu, Z.H. [Shanghai Nuclear Engineering Research and Design Institute, Shanghai (China); Yang, Y.H. [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, 800 Dong Chuan Road, Shanghai 200240 (China); Cheng, X. [Institute of Fusion and Nuclear Technology, Karlsruhe Institute of Technology (KIT), Kaiserstr. 12, 76131 Karlsruhe (Germany)

    2013-06-15

    Highlights: • Application of the ATHLET-SC code to the trans-critical analysis for SCWR. • Development of a passive safety system for SCWR-M. • Analysis of hot/cold leg LOCA behaviour with different break size. • Introduction of some mitigation measures for SCWR-M -- Abstract: A new SCWR conceptual design (mixed spectrum supercritical water cooled reactor: SCWR-M) is proposed by Shanghai Jiao Tong University (SJTU). R and D activities covering core design, safety system design and code development of SCWR-M are launched at SJTU. Safety system design and analysis is one of the key tasks during the development of SCWR-M. Considering the current advanced reactor design, a new passive safety system for SCWR-M including isolation cooling system (ICS), accumulator injection system (ACC), gravity driven cooling system (GDCS) and automatic depressurization system (ADS) is proposed. Based on the modified and preliminarily assessed system code ATHLET-SC, loss of coolant accident (LOCA) analysis for hot and cold leg is performed in this paper. Three different break sizes are analyzed to clarify the hot and cold LOCA characteristics of the SCWR-M. The influence of the break location and break size on the safety performance of SCWR-M is also concluded. Several measures to induce the core coolant flow and to mitigate core heating up are also discussed. The results achieved so far demonstrate the feasibility of the proposed passive safety system to keep the SCWR-M core at safety condition during loss of coolant accident.

  17. PA activity by using nuclear power plant safety demonstration and analysis

    International Nuclear Information System (INIS)

    Tsuchiya, Mitsuo; Kamimae, Rie

    1999-01-01

    INS/NUPEC presents one of Public acceptance (PA) methods for nuclear power in Japan, 'PA activity by using Nuclear Power Plant Safety Demonstration and Analysis', by using one of videos which is explained and analyzed accident events (Loss of Coolant Accident). Safety regulations of The National Government are strictly implemented in licensing at each of basic design and detailed design. To support safety regulation activities conducted by the National Government, INS/NLTPEC continuously implement Safety demonstration and analysis. With safety demonstration and analysis, made by assuming some abnormal conditions, what impacts could be produced by the assumed conditions are forecast based on specific design data on a given nuclear power plants. When analysis results compared with relevant decision criteria, the safety of nuclear power plants is confirmed. The decision criteria are designed to help judge if or not safety design of nuclear power plants is properly made. The decision criteria are set in the safety examination guidelines by taking sufficient safety allowance based on the latest technical knowledge obtained from a wide range of tests and safety studies. Safety demonstration and analysis is made by taking the procedure which are summarized in this presentation. In Japan, various PA (Public Acceptance) pamphlets and videos on nuclear energy have been published. But many of them focused on such topics as necessity or importance of nuclear energy, basic principles of nuclear power generation, etc., and a few described safety evaluation particularly of abnormal and accident events in accordance with the regulatory requirements. In this background, INS/NUPEC has been making efforts to prepare PA pamphlets and videos to explain the safety of nuclear power plants, to be simple and concrete enough, using various analytical computations for abnormal and accident events. In results, PA activity of INS/NUPEC is evaluated highly by the people

  18. Waste Isolation Pilot Plant Safety Analysis Report

    International Nuclear Information System (INIS)

    1995-11-01

    The following provides a summary of the specific issues addressed in this FY-95 Annual Update as they relate to the CH TRU safety bases: Executive Summary; Site Characteristics; Principal Design and Safety Criteria; Facility Design and Operation; Hazards and Accident Analysis; Derivation of Technical Safety Requirements; Radiological and Hazardous Material Protection; Institutional Programs; Quality Assurance; and Decontamination and Decommissioning. The System Design Descriptions'' (SDDS) for the WIPP were reviewed and incorporated into Chapter 3, Principal Design and Safety Criteria and Chapter 4, Facility Design and Operation. This provides the most currently available final engineering design information on waste emplacement operations throughout the disposal phase up to the point of permanent closure. Also, the criteria which define the TRU waste to be accepted for disposal at the WIPP facility were summarized in Chapter 3 based on the WAC for the Waste Isolation Pilot Plant.'' This Safety Analysis Report (SAR) documents the safety analyses that develop and evaluate the adequacy of the Waste Isolation Pilot Plant Contact-Handled Transuranic Wastes (WIPP CH TRU) safety bases necessary to ensure the safety of workers, the public and the environment from the hazards posed by WIPP waste handling and emplacement operations during the disposal phase and hazards associated with the decommissioning and decontamination phase. The analyses of the hazards associated with the long-term (10,000 year) disposal of TRU and TRU mixed waste, and demonstration of compliance with the requirements of 40 CFR 191, Subpart B and 40 CFR 268.6 will be addressed in detail in the WIPP Final Certification Application scheduled for submittal in October 1996 (40 CFR 191) and the No-Migration Variance Petition (40 CFR 268.6) scheduled for submittal in June 1996. Section 5.4, Long-Term Waste Isolation Assessment summarizes the current status of the assessment

  19. Waste Isolation Pilot Plant Safety Analysis Report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-11-01

    The following provides a summary of the specific issues addressed in this FY-95 Annual Update as they relate to the CH TRU safety bases: Executive Summary; Site Characteristics; Principal Design and Safety Criteria; Facility Design and Operation; Hazards and Accident Analysis; Derivation of Technical Safety Requirements; Radiological and Hazardous Material Protection; Institutional Programs; Quality Assurance; and Decontamination and Decommissioning. The System Design Descriptions`` (SDDS) for the WIPP were reviewed and incorporated into Chapter 3, Principal Design and Safety Criteria and Chapter 4, Facility Design and Operation. This provides the most currently available final engineering design information on waste emplacement operations throughout the disposal phase up to the point of permanent closure. Also, the criteria which define the TRU waste to be accepted for disposal at the WIPP facility were summarized in Chapter 3 based on the WAC for the Waste Isolation Pilot Plant.`` This Safety Analysis Report (SAR) documents the safety analyses that develop and evaluate the adequacy of the Waste Isolation Pilot Plant Contact-Handled Transuranic Wastes (WIPP CH TRU) safety bases necessary to ensure the safety of workers, the public and the environment from the hazards posed by WIPP waste handling and emplacement operations during the disposal phase and hazards associated with the decommissioning and decontamination phase. The analyses of the hazards associated with the long-term (10,000 year) disposal of TRU and TRU mixed waste, and demonstration of compliance with the requirements of 40 CFR 191, Subpart B and 40 CFR 268.6 will be addressed in detail in the WIPP Final Certification Application scheduled for submittal in October 1996 (40 CFR 191) and the No-Migration Variance Petition (40 CFR 268.6) scheduled for submittal in June 1996. Section 5.4, Long-Term Waste Isolation Assessment summarizes the current status of the assessment.

  20. Posttest analysis of the FFTF inherent safety tests

    International Nuclear Information System (INIS)

    Padilla, A. Jr.; Claybrook, S.W.

    1987-01-01

    Inherent safety tests were performed during 1986 in the 400-MW (thermal) Fast Flux Test Facility (FFTF) reactor to demonstrate the effectiveness of an inherent shutdown device called the gas expansion module (GEM). The GEM device provided a strong negative reactivity feedback during loss-of-flow conditions by increasing the neutron leakage as a result of an expanding gas bubble. The best-estimate pretest calculations for these tests were performed using the IANUS plant analysis code (Westinghouse Electric Corporation proprietary code) and the MELT/SIEX3 core analysis code. These two codes were also used to perform the required operational safety analyses for the FFTF reactor and plant. Although it was intended to also use the SASSYS systems (core and plant) analysis code, the calibration of the SASSYS code for FFTF core and plant analysis was not completed in time to perform pretest analyses. The purpose of this paper is to present the results of the posttest analysis of the 1986 FFTF inherent safety tests using the SASSYS code

  1. SNF fuel retrieval sub project safety analysis document

    International Nuclear Information System (INIS)

    BERGMANN, D.W.

    1999-01-01

    This safety analysis is for the SNF Fuel Retrieval (FRS) Sub Project. The FRS equipment will be added to K West and K East Basins to facilitate retrieval, cleaning and repackaging the spent nuclear fuel into Multi-Canister Overpack baskets. The document includes a hazard evaluation, identifies bounding accidents, documents analyses of the accidents and establishes safety class or safety significant equipment to mitigate accidents as needed

  2. SNF fuel retrieval sub project safety analysis document

    Energy Technology Data Exchange (ETDEWEB)

    BERGMANN, D.W.

    1999-02-24

    This safety analysis is for the SNF Fuel Retrieval (FRS) Sub Project. The FRS equipment will be added to K West and K East Basins to facilitate retrieval, cleaning and repackaging the spent nuclear fuel into Multi-Canister Overpack baskets. The document includes a hazard evaluation, identifies bounding accidents, documents analyses of the accidents and establishes safety class or safety significant equipment to mitigate accidents as needed.

  3. Development and improvement of safety analysis code for geological disposal

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    In order to confirm the long-term safety concerning geological disposal, probabilistic safety assessment code and other analysis codes, which can evaluate possibility of each event and influence on engineered barrier and natural barrier by the event, were introduced. We confirmed basic functions of those codes and studied the relation between those functions and FEP/PID which should be taken into consideration in safety assessment. We are planning to develop 'Nuclide Migration Assessment System' for the purpose of realizing improvement in efficiency of assessment work, human error prevention for analysis, and quality assurance of the analysis environment and analysis work for safety assessment by using it. As the first step, we defined the system requirements and decided the system composition and functions which should be mounted in them based on those requirements. (author)

  4. Safety analysis of control rod drive computers

    International Nuclear Information System (INIS)

    Ehrenberger, W.; Rauch, G.; Schmeil, U.; Maertz, J.; Mainka, E.U.; Nordland, O.; Gloee, G.

    1985-01-01

    The analysis of the most significant user programmes revealed no errors in these programmes. The evaluation of approximately 82 cumulated years of operation demonstrated that the operating system of the control rod positioning processor has a reliability that is sufficiently good for the tasks this computer has to fulfil. Computers can be used for safety relevant tasks. The experience gained with the control rod positioning processor confirms that computers are not less reliable than conventional instrumentation and control system for comparable tasks. The examination and evaluation of computers for safety relevant tasks can be done with programme analysis or statistical evaluation of the operating experience. Programme analysis is recommended for seldom used and well structured programmes. For programmes with a long, cumulated operating time a statistical evaluation is more advisable. The effort for examination and evaluation is not greater than the corresponding effort for conventional instrumentation and control systems. This project has also revealed that, where it is technologically sensible, process controlling computers or microprocessors can be qualified for safety relevant tasks without undue effort. (orig./HP) [de

  5. Safety evaluation report of the Waste Isolation Pilot Plant safety analysis report: Contact-handled transuranic waste disposal operations

    International Nuclear Information System (INIS)

    1997-02-01

    DOE 5480.23, Nuclear Safety Analysis Reports, requires that the US Department of Energy conduct an independent, defensible, review in order to approve a Safety Analysis Report (SAR). That review and the SAR approval basis is documented in this formal Safety Evaluation Report (SER). This SER documents the DOE's review of the Waste Isolation Pilot Plant SAR and provides the Carlsbad Area Office Manager, the WIPP SAR approval authority, with the basis for approving the safety document. It concludes that the safety basis documented in the WIPP SAR is comprehensive, correct, and commensurate with hazards associated with planned waste disposal operations

  6. FAST and SAFE Passive Safety Devices for Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hartanto, Donny; Kim, Chihyung; Kim, In-Hyung; Kim, Yonghee [KAIST, Daejeon (Korea, Republic of)

    2015-05-15

    The major factor is the impact of the neutron spectral hardening. The second factor that affects the CVR is reduced capture by the coolant when the coolant voiding occurs. To improve the CVR, many ideas and concepts have been proposed, which include introduction of an internal blanket, spectrum softening, or increasing the neutron leakage. These ideas may reduce the CVR, but they deteriorate the neutron economy. Another potential solution is to adopt a passive safety injection device such as the ARC (autonomous reactivity control) system, which is still under development. In this paper, two new concepts of passive safety devices are proposed. The devices are called FAST (Floating Absorber for Safety at Transient) and SAFE (Static Absorber Feedback Equipment). Their purpose is to enhance the negative reactivity feedback originating from the coolant in fast reactors. SAFE is derived to balance the positive reactivity feedback due to sodium coolant temperature increases. It has been demonstrated that SAFE allows a low-leakage SFR to achieve a self-shutdown and self-controllability even though the generic coolant temperature coefficient is quite positive and the coolant void reactivity can be largely managed by the new FAST device. It is concluded that both FAST and SAFE devices will improve substantially the fast reactor safety and they deserve more detailed investigations.

  7. Safety analysis of the existing 850 Firing Facility

    International Nuclear Information System (INIS)

    Odell, B.N.

    1986-01-01

    A safety analysis was performed to determine if normal operations and/or potential accidents at the 850 Firing Facility at Site 300 could present undue hazards to the general public, personnel at Site 300, or have an adverse effect on the environment. The normal operations and credible accidents that might have an effect on these facilities or have off-site consequences were considered. It was determined by this analysis that all but one of the hazards were either low or of the type or magnitude routinely encountered and/or accepted by the public. The exception was explosives, which was classified as a moderate hazard per the requirements given in DOE Order 5481.1A. This safety analysis concluded that the operation at this facility will present no undue risk to the health and safety of LLNL employees or the public

  8. Safety analysis of the existing 851 Firing Facility

    International Nuclear Information System (INIS)

    Odell, B.N.

    1986-01-01

    A safety analysis was performed to determine if normal operations and/or potential accidents at the 851 Firing Facility at Site 300 could present undue hazards to the general public, personnel at Site 300, or have an adverse effect on the environment. The normal operations and credible accidents that might have an effect on these facilities or have off-site consequences were considered. It was determined by this analysis that all but two of the hazards were either low or of the type or magnitude routinely encountered and/or accepted by the public. The exceptions were the linear accelerator and explosives, which were classified as moderate hazards per the requirements given in DOE Order 5481.1A. This safety analysis concluded that the operation at this facility will present no undue risk to the health and safety of LLNL employees or the public

  9. Statistical margin to DNB safety analysis approach for LOFT

    International Nuclear Information System (INIS)

    Atkinson, S.A.

    1982-01-01

    A method was developed and used for LOFT thermal safety analysis to estimate the statistical margin to DNB for the hot rod, and to base safety analysis on desired DNB probability limits. This method is an advanced approach using response surface analysis methods, a very efficient experimental design, and a 2nd-order response surface equation with a 2nd-order error propagation analysis to define the MDNBR probability density function. Calculations for limiting transients were used in the response surface analysis thereby including transient interactions and trip uncertainties in the MDNBR probability density

  10. Tolerability of risk, safety assessment principles and their implications for probabilistic safety analysis

    International Nuclear Information System (INIS)

    Ewing, D.J.F.; Campbell, J.F.

    1994-01-01

    This paper gives a regulatory view of probabilistic safety assessment as seen by the Nuclear Installations Inspectorate (NII) and in the light of the general regulatory risk aims set out in the Health and Safety Executive's (HSE) The tolerability of risk from nuclear power stations (TOR) and in Safety assessment principles for nuclear plants (SAPs), prepared by NII on behalf of the HSE. Both of these publications were revised and republished in 1992. This paper describes the SAPs, together with the historical background, the motivation for review, the effects of the Sizewell and Hinkley Point C public inquiries, changes since the original versions, comparison with international standards and use in assessment. For new plant, probabilistic safety analysis (PSA) is seen as an essential tool in balancing the safety of the design and in demonstrating compliance with TOR and the SAPs. (Author)

  11. Development of regulatory technology for thermal-hydraulic safety analysis

    International Nuclear Information System (INIS)

    Bang, Young Seok; Lee, S. H.; Ryu, Y. H.

    2001-02-01

    The present study aims to develop the regulation capability in thermal-hydraulic safety analysis which was required for the reasonable safety regulation in the current NPP, the next generation reactors, and the future-type reactors. The fourth fiscal year of the first phase of the research was focused on the following research topics: Investigation on the current status of the thermal-hydraulic safety analysis technology outside and inside of the country; Review on the improved features of the thermal-hydraulic safety analysis regulatory audit code, RELAP5/MOD3; Assessments of code with LOFT L9-3 ATWS experiment and LSTF SB-SG-10 multiple SGTR experiment; Application of the RELAP5/CANDU code to analyses of SLB and LBLOCA and evaluation of its effect on safety; Application of the code to IAEA PHWR ISP analysis; Assessments of RELAP5 and TRAC with UPTF downcomer injection test and Analysis of LBLOCA with RELAP5 for the performance evaluation of KNGR DVI; Setup of a coupled 3-D kinetics and thermal-hydraulics and application it to a reactivity accident analysis; and Extension of database and improvement of plant input decks. For supporting the resolution of safety issues, loss of RHR event during midloop operation was analyzed for Kori Unit 3, issues on high burnup fuel were reviewed and performance of FRAPCON-3 assessed. Also MSLB was analyzed to figure out the sensitivity of downcomer temperature supporting the PTS risk evaluation of Kori Unit 1. Thermal stratification in pipe was analyzed using the method proposed. And a method predicting the thermal-hydraulic performance of IRWST of KNGR was explored. The PWR ECCS performance criteria was issued as a MOST Article 200-19.and a regulatory guide on evaluation methodology was improved to cover concerns raised from the related licensing review process

  12. Safety analysis in support of regulatory decision marking

    International Nuclear Information System (INIS)

    Pomier Baez, L.; Troncoso Fleitas, M.; Valhuerdi Debesa, C.; Valle Cepero, R.; Hernandez, J.L.

    1996-01-01

    Features of different safety analysis techniques by means of calculation thermohydraulic a probabilistic and severe accidents used in the safety assessment, as well as the development of these techniques in Cuba and their use in support of regulatory decision making are presented

  13. Operational safety analysis status of Novi Han repository

    International Nuclear Information System (INIS)

    Boiadjiev, A.

    2000-01-01

    This article presents the status of the safety studies and activities related to Novi Han repository. The case of this facility is such that no clear boundary exists between post-closure safety assessment and operational safety assessment. The major findings of these activities are given. The Safety Analysis Report (SAR) for Novi Han repository is developed by Risk Engineering Ltd. under a contract with the Committee on the Use of Atomic Energy for Peaceful Purposes. The general structure and main conclusions and recommendations of the SAR are presented. (author)

  14. Safety analysis and evaluation methodology for fusion systems

    International Nuclear Information System (INIS)

    Fujii-e, Y.; Kozawa, Y.; Namba, C.

    1987-03-01

    Fusion systems which are under development as future energy systems have reached a stage that the break even is expected to be realized in the near future. It is desirable to demonstrate that fusion systems are well acceptable to the societal environment. There are three crucial viewpoints to measure the acceptability, that is, technological feasibility, economy and safety. These three points have close interrelation. The safety problem is more important since three large scale tokamaks, JET, TFTR and JT-60, start experiment, and tritium will be introduced into some of them as the fusion fuel. It is desirable to establish a methodology to resolve the safety-related issues in harmony with the technological evolution. The promising fusion system toward reactors is not yet settled. This study has the objective to develop and adequate methodology which promotes the safety design of general fusion systems and to present a basis for proposing the R and D themes and establishing the data base. A framework of the methodology, the understanding and modeling of fusion systems, the principle of ensuring safety, the safety analysis based on the function and the application of the methodology are discussed. As the result of this study, the methodology for the safety analysis and evaluation of fusion systems was developed. New idea and approach were presented in the course of the methodology development. (Kako, I.)

  15. Gap Analysis Approach for Construction Safety Program Improvement

    Directory of Open Access Journals (Sweden)

    Thanet Aksorn

    2007-06-01

    Full Text Available To improve construction site safety, emphasis has been placed on the implementation of safety programs. In order to successfully gain from safety programs, factors that affect their improvement need to be studied. Sixteen critical success factors of safety programs were identified from safety literature, and these were validated by safety experts. This study was undertaken by surveying 70 respondents from medium- and large-scale construction projects. It explored the importance and the actual status of critical success factors (CSFs. Gap analysis was used to examine the differences between the importance of these CSFs and their actual status. This study found that the most critical problems characterized by the largest gaps were management support, appropriate supervision, sufficient resource allocation, teamwork, and effective enforcement. Raising these priority factors to satisfactory levels would lead to successful safety programs, thereby minimizing accidents.

  16. Multi-dimensional approach of MARS-LMR for the analysis of Phenix End-of-Life natural circulation test

    International Nuclear Information System (INIS)

    Jeong, Hae Yong; Ha, Kwi Seok; Chang, Won Pyo; Lee, Kwi Lim

    2012-01-01

    Phenix is one of the important prototype sodium-cooled fast reactors (SFR) in nuclear reactor development history. It had been operated successfully for 35 years by the French Commissariat a l'energie atomique (CEA) and the Electricite de France (EdF) achieving its original objectives of demonstrating a fast breeder reactor technology and of playing the role of irradiation facility for innovative fuels and materials. After its final shutdown in 2009, CEA launched the Phenix End-of-life (EOL) test program. It provided a unique opportunity to generate reliable test data which is inevitable in the validation and verification of a SFR system analysis code. KAERI joined this international collaboration program of IAEA CRP and has performed the pretest analysis and post-test analysis utilizing the one-dimensional modeling of the MARS-LMR code, which had been developed by KAERI for the transient analysis of SFR systems. Through the previous studies, it has been identified that there are some limitations in the modeling of complicated thermal-hydraulic behaviors in the large pool volumes with the one-dimensional modeling. Recently, KAERI performed the analysis of Phenix EOL natural circulation test with multi-dimensional pool modeling, which is detailed below

  17. Multi-dimensional approach of MARS-LMR for the analysis of Phenix End-of-Life natural circulation test

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Hae Yong; Ha, Kwi Seok; Chang, Won Pyo; Lee, Kwi Lim [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-05-15

    Phenix is one of the important prototype sodium-cooled fast reactors (SFR) in nuclear reactor development history. It had been operated successfully for 35 years by the French Commissariat a l'energie atomique (CEA) and the Electricite de France (EdF) achieving its original objectives of demonstrating a fast breeder reactor technology and of playing the role of irradiation facility for innovative fuels and materials. After its final shutdown in 2009, CEA launched the Phenix End-of-life (EOL) test program. It provided a unique opportunity to generate reliable test data which is inevitable in the validation and verification of a SFR system analysis code. KAERI joined this international collaboration program of IAEA CRP and has performed the pretest analysis and post-test analysis utilizing the one-dimensional modeling of the MARS-LMR code, which had been developed by KAERI for the transient analysis of SFR systems. Through the previous studies, it has been identified that there are some limitations in the modeling of complicated thermal-hydraulic behaviors in the large pool volumes with the one-dimensional modeling. Recently, KAERI performed the analysis of Phenix EOL natural circulation test with multi-dimensional pool modeling, which is detailed below

  18. Incorporating Traffic Control and Safety Hardware Performance Functions into Risk-based Highway Safety Analysis

    Directory of Open Access Journals (Sweden)

    Zongzhi Li

    2017-04-01

    Full Text Available Traffic control and safety hardware such as traffic signs, lighting, signals, pavement markings, guardrails, barriers, and crash cushions form an important and inseparable part of highway infrastructure affecting safety performance. Significant progress has been made in recent decades to develop safety performance functions and crash modification factors for site-specific crash predictions. However, the existing models and methods lack rigorous treatments of safety impacts of time-deteriorating conditions of traffic control and safety hardware. This study introduces a refined method for computing the Safety Index (SI as a means of crash predictions for a highway segment that incorporates traffic control and safety hardware performance functions into the analysis. The proposed method is applied in a computation experiment using five-year data on nearly two hundred rural and urban highway segments. The root-mean square error (RMSE, Chi-square, Spearman’s rank correlation, and Mann-Whitney U tests are employed for validation.

  19. Quantitative Safety and Security Analysis from a Communication Perspective

    Directory of Open Access Journals (Sweden)

    Boris Malinowsky

    2015-12-01

    Full Text Available This paper introduces and exemplifies a trade-off analysis of safety and security properties in distributed systems. The aim is to support analysis for real-time communication and authentication building blocks in a wireless communication scenario. By embedding an authentication scheme into a real-time communication protocol for safety-critical scenarios, we can rely on the protocol’s individual safety and security properties. The resulting communication protocol satisfies selected safety and security properties for deployment in safety-critical use-case scenarios with security requirements. We look at handover situations in a IEEE 802.11 wireless setup between mobile nodes and access points. The trade-offs involve application-layer data goodput, probability of completed handovers, and effect on usable protocol slots, to quantify the impact of security from a lower-layer communication perspective on the communication protocols. The results are obtained using the network simulator ns-3.

  20. Safety of radiation sources in Slovenia

    International Nuclear Information System (INIS)

    Belicic-Kolsek, A.; Sutej, T.

    2001-01-01

    The Republic of Slovenia, a central European country which has been independent since 1991, has about 2 million inhabitants and an area of 20,256 km 2 . The Constitutional Law on Enforcement of the Basic Constitutional Charter on the Autonomy and Independence of the Republic of Slovenia, adopted on 23 June 1991 (Off. Gaz. of the R of Slovenia No. 1/91), provided that all the laws adopted by the Socialist Federal Republic (SFR) of Yugoslavia should remain in force in the Republic of Slovenia pending the adoption of appropriate legislation by the Slovene Parliament. Under the Slovene Constitution, all international treaties ratified by Slovenia constitute an integral part of Slovenia's legislation and can be applied directly. In Slovenia, all regular types of ionizing radiation source are being used for peaceful purposes and are covered by a system for their safe use and control. All radiation sources and radioactive materials are registered and under regulatory control. Inspections are carried out periodically by the Health Inspectorate of the Republic of Slovenia (HIRS) and, in the case of nuclear installations, the Slovene Nuclear Safety Administration (SNSA). Technical checks on radiation sources are carried out periodically by technical support organizations: the Jozef Stefan Institute and the Institute for Occupational Safety (IOS). (author)

  1. Safety analysis and review system: a Department of Energy safety assurance tool

    International Nuclear Information System (INIS)

    Rosenthal, H.B.

    1981-01-01

    The concept of the Safety Analysis and Review System is not new. It has been used within the Department and its predecessor agencies, Atomic Energy Commission (AEC) and Energy Research and Development Administration (ERDA), for over 20 years. To minimize the risks from nuclear reactor and power plants, the AEC developed a process to support management authorization of each operation through identification and analysis of potential hazards and the measures taken to control them. As the agency evolved from AEC through ERDA to the Department of Energy, its responsibilities were broadened to cover a diversity of technologies, including those associated with the development of fossil, solar, and geothermal energy. Because the safety analysis process had proved effective in a technology of high potential hazard, the Department investigated the applicability of the process to the other technologies. This paper describes the system and discusses how it is implemented within the Department

  2. Upgrading the safety toolkit: Initiatives of the accident analysis subgroup

    International Nuclear Information System (INIS)

    O'Kula, K.R.; Chung, D.Y.

    1999-01-01

    Since its inception, the Accident Analysis Subgroup (AAS) of the Energy Facility Contractors Group (EFCOG) has been a leading organization promoting development and application of appropriate methodologies for safety analysis of US Department of Energy (DOE) installations. The AAS, one of seven chartered by the EFCOG Safety Analysis Working Group, has performed an oversight function and provided direction to several technical groups. These efforts have been instrumental toward formal evaluation of computer models, improving the pedigree on high-use computer models, and development of the user-friendly Accident Analysis Guidebook (AAG). All of these improvements have improved the analytical toolkit for best complying with DOE orders and standards shaping safety analysis reports (SARs) and related documentation. Major support for these objectives has been through DOE/DP-45

  3. Preliminary Studies of Two-Phase Reactive Process of Sodium-CO2 in S-CO2 Power Conversion Cycle Coupled to SFR System

    International Nuclear Information System (INIS)

    Jung, Hwa Young; Ahn, Yoon Han; Lee, You Ho; Lee, Jeong Ik

    2013-01-01

    As a competing alternative to the steam Rankine cycle, the supercritical CO 2 (S-CO 2 ) Brayton cycle has been highlighted due to its high thermal efficiency, compact turbomachinery and heat exchangers sizes, and the reduced risk of SWRs. While the reduced risk of an SWR is considered as the one of most pronounced benefits of S-CO 2 Brayton cycle, there is still an interaction problem between liquid sodium and CO 2 . Although the chemical interaction between liquid sodium and CO 2 demonstrates less serious potential risks than those of a SWR, the Na/CO 2 interaction should be understood to evaluate safety and reliability of Intermediate Heat eXchanger (IHX). A noticeable characteristic of the reaction environment is that there is a large pressure difference between the liquid sodium and CO 2 side by about 1 and 200 bar, respectively. This would imply that the presence of a micro-crack in a heat exchanger tube will cause a high-pressure leak of CO 2 into liquid sodium side. Although the Na/CO 2 interaction may play an important role in the safety of the SFR reactor system, there has not yet been any research on understanding Na/CO 2 reaction by leakage through IHX. For this problem, the Korea Advanced Institute of Science and Technology (KAIST) research team is studying the mechanism of CO 2 leakage and Na/CO 2 interaction in more details. The KAIST research team developed the MATLAB code, KAIST H XD, which can be used to design and evaluate performance of a heat exchanger of an S-CO 2 cycle. The size of heat exchanger and the amount of CO 2 in the cycle are calculated from the KAIST H XD code to estimate the amount of reaction products in Na/CO 2 interaction as well as liquid sodium

  4. 3D analysis methods - Study and seminar[BWR safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Daaviittila, A [Valtion Teknillinen Tutkimuskeskus (Finland)

    2003-10-01

    The first part of the report results from a study that was performed as a Nordic co-operation activity with active participation from Studsvik Scandpower and Westinghouse Atom in Sweden, and VTT in Finland. The purpose of the study was to identify and investigate the effects rising from using the 3D transient com-puter codes in BWR safety analysis, and their influence on the transient analysis methodology. One of the main questions involves the critical power ratio (CPR) calculation methodology. The present way, where the CPR calculation is per-formed with a separate hot channel calculation, can be artificially conservative. In the investigated cases, no dramatic minimum CPR effect coming from the 3D calculation is apparent. Some cases show some decrease in the transient change of minimum CPR with the 3D calculation, which confirms the general thinking that the 1D calculation is conservative. On the other hand, the observed effect on neutron flux behaviour is quite large. In a slower transient the 3D effect might be stronger. The second part of the report is a summary of a related seminar that was held on the 3D analysis methods. The seminar was sponsored by the Reactor Safety part (NKS-R) of the Nordic Nuclear Safety Research Programme (NKS). (au)

  5. Analysis approach for common cause failure on non-safety digital control system

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yun Goo; Oh, Eungse [Korea Hydro and Nuclear Power Co. Ltd., Daejeon (Korea, Republic of)

    2014-05-15

    The effects of common cause failure (CCF) on safety digital instrumentation and control (I and C) system had been considered in defense in depth and diversity coping analysis with safety analysis method. For the non-safety system, single failure had been considered for safety analysis. IEEE Std. 603-1991, Clause 5.6.3.1(2), 'Isolation' states that no credible failure on the non-safety side of an isolation device shall prevent any portion of a safety system from meeting its minimum performance requirements during and following any design basis event requiring that safety function. The software CCF is one of the credible failure on the non-safety side. In advanced digital I and C system, same hardware component is used for different control system and the defect in manufacture or common external event can generate CCF. Moreover, the non-safety I and C system uses complex software for its various function and software quality assurance for the development process is less severe than safety software for the cost effective design. Therefore the potential defects in software cannot be ignored and the effect of software CCF on non-safety I and C system is needed to be evaluated. This paper proposes the general process and considerations for the analysis of CCF on non-safety I and C system.

  6. Safety analysis and synthesis using fuzzy sets and evidential reasoning

    International Nuclear Information System (INIS)

    Wang, J.; Yang, J.B.; Sen, P.

    1995-01-01

    This paper presents a new methodology for safety analysis and synthesis of a complex engineering system with a structure that is capable of being decomposed into a hierarchy of levels. In this methodology, fuzzy set theory is used to describe each failure event and an evidential reasoning approach is then employed to synthesise the information thus produced to assess the safety of the whole system. Three basic parameters--failure likelihood, consequence severity and failure consequence probability, are used to analyse a failure event. These three parameters are described by linguistic variables which are characterised by a membership function to the defined categories. As safety can also be clearly described by linguistic variables referred to as the safety expressions, the obtained fuzzy safety score can be mapped back to the safety expressions which are characterised by membership functions over the same categories. This mapping results in the identification of the safety of each failure event in terms of the degree to which the fuzzy safety score belongs to each of the safety expressions. Such degrees represent the uncertainty in safety evaluations and can be synthesised using an evidential reasoning approach so that the safety of the whole system can be evaluated in terms of these safety expressions. Finally, a practical engineering example is presented to demonstrate the proposed safety analysis and synthesis methodology

  7. Bayesian-network-based safety risk analysis in construction projects

    International Nuclear Information System (INIS)

    Zhang, Limao; Wu, Xianguo; Skibniewski, Miroslaw J.; Zhong, Jingbing; Lu, Yujie

    2014-01-01

    This paper presents a systemic decision support approach for safety risk analysis under uncertainty in tunnel construction. Fuzzy Bayesian Networks (FBN) is used to investigate causal relationships between tunnel-induced damage and its influential variables based upon the risk/hazard mechanism analysis. Aiming to overcome limitations on the current probability estimation, an expert confidence indicator is proposed to ensure the reliability of the surveyed data for fuzzy probability assessment of basic risk factors. A detailed fuzzy-based inference procedure is developed, which has a capacity of implementing deductive reasoning, sensitivity analysis and abductive reasoning. The “3σ criterion” is adopted to calculate the characteristic values of a triangular fuzzy number in the probability fuzzification process, and the α-weighted valuation method is adopted for defuzzification. The construction safety analysis progress is extended to the entire life cycle of risk-prone events, including the pre-accident, during-construction continuous and post-accident control. A typical hazard concerning the tunnel leakage in the construction of Wuhan Yangtze Metro Tunnel in China is presented as a case study, in order to verify the applicability of the proposed approach. The results demonstrate the feasibility of the proposed approach and its application potential. A comparison of advantages and disadvantages between FBN and fuzzy fault tree analysis (FFTA) as risk analysis tools is also conducted. The proposed approach can be used to provide guidelines for safety analysis and management in construction projects, and thus increase the likelihood of a successful project in a complex environment. - Highlights: • A systemic Bayesian network based approach for safety risk analysis is developed. • An expert confidence indicator for probability fuzzification is proposed. • Safety risk analysis progress is extended to entire life cycle of risk-prone events. • A typical

  8. Galileo and Ulysses missions safety analysis and launch readiness status

    International Nuclear Information System (INIS)

    Cork, M.J.; Turi, J.A.

    1989-01-01

    The Galileo spacecraft will explore the Jupiter system and Ulysses will fly by Jupiter en route to a polar orbit of the sun. Both spacecraft are powered by general purpose heat source radioisotope thermoelectric generators (RTGs). As a result of the Challenger accident and subsequent mission reprogramming, the Galileo and Ulysses missions' safety analysis had to be repeated. In addition to presenting an overview of the safety analysis status for the missions, this paper presents a brief review of the missions' objectives and design approaches, RTG design characteristics and development history, and a description of the safety analysis process. (author)

  9. A root cause analysis project in a medication safety course.

    Science.gov (United States)

    Schafer, Jason J

    2012-08-10

    To develop, implement, and evaluate team-based root cause analysis projects as part of a required medication safety course for second-year pharmacy students. Lectures, in-class activities, and out-of-class reading assignments were used to develop students' medication safety skills and introduce them to the culture of medication safety. Students applied these skills within teams by evaluating cases of medication errors using root cause analyses. Teams also developed error prevention strategies and formally presented their findings. Student performance was assessed using a medication errors evaluation rubric. Of the 211 students who completed the course, the majority performed well on root cause analysis assignments and rated them favorably on course evaluations. Medication error evaluation and prevention was successfully introduced in a medication safety course using team-based root cause analysis projects.

  10. Establishment of joint application system of safety analysis codes between Korea and Vietnam

    International Nuclear Information System (INIS)

    Chung, Bub Dong; Kim, Kyung Doo; Park, Cheol; Bae, Sung Won; Baek, Won Pil; Song, Cheol hwa; Jeong, Jae Jun; Lee, Seung Wook; Hwang, Moon Kyu; Lee, Chang Sup

    2011-04-01

    The following KAERI-VAEI collaboration works have been performed during the 2 year project ('09.4∼'11.4). 1) On the job training of Vietnam code users(1st training for 4 VAEI staff-3 months. 2nd training for 3 VAEI staff- 3 month), 2) Lecture of nuclear safety analysis (30 hrs basic course and 30 hrs advanced course), 3) Review of safety analysis method (IAEA safety concept and requirements), 4) Collaborative assessment of safety analysis code MARS (13 conceptual problem, 2 separate effect test problem, 1 integral effect test problem), 5) Input deck preparation of standard PWR (Preparation of APR1400 input deck and safety analysis of DBA). VAEI staffs have been familiarized to Korean PWR safety assessment technology through the collaboration assessment work using a computer code developed in Korea. The lectures for Vietnamese research will be contributed to the utilization and cultivation of Korean safety technology. The collaborated assessment works will be used for the establishment of MARS based safety analysis system which is independent from US safety assessment system

  11. Establishment of joint application system of safety analysis codes between Korea and Vietnam

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Bub Dong; Kim, Kyung Doo; Park, Cheol; Bae, Sung Won; Baek, Won Pil; Song, Cheol hwa; Jeong, Jae Jun; Lee, Seung Wook; Hwang, Moon Kyu; Lee, Chang Sup [KAERI, Daejeon (Korea, Republic of)

    2011-04-15

    The following KAERI-VAEI collaboration works have been performed during the 2 year project ('09.4{approx}'11.4). 1) On the job training of Vietnam code users(1st training for 4 VAEI staff-3 months. 2nd training for 3 VAEI staff- 3 month), 2) Lecture of nuclear safety analysis (30 hrs basic course and 30 hrs advanced course), 3) Review of safety analysis method (IAEA safety concept and requirements), 4) Collaborative assessment of safety analysis code MARS (13 conceptual problem, 2 separate effect test problem, 1 integral effect test problem), 5) Input deck preparation of standard PWR (Preparation of APR1400 input deck and safety analysis of DBA). VAEI staffs have been familiarized to Korean PWR safety assessment technology through the collaboration assessment work using a computer code developed in Korea. The lectures for Vietnamese research will be contributed to the utilization and cultivation of Korean safety technology. The collaborated assessment works will be used for the establishment of MARS based safety analysis system which is independent from US safety assessment system

  12. Reliability analysis of Angra I safety systems

    International Nuclear Information System (INIS)

    Oliveira, L.F.S. de; Soto, J.B.; Maciel, C.C.; Gibelli, S.M.O.; Fleming, P.V.; Arrieta, L.A.

    1980-07-01

    An extensive reliability analysis of some safety systems of Angra I, are presented. The fault tree technique, which has been successfully used in most reliability studies of nuclear safety systems performed to date is employed. Results of a quantitative determination of the unvailability of the accumulator and the containment spray injection systems are presented. These results are also compared to those reported in WASH-1400. (E.G.) [pt

  13. Safety analysis methodologies for radioactive waste repositories in shallow ground

    International Nuclear Information System (INIS)

    1984-01-01

    The report is part of the IAEA Safety Series and is addressed to authorities and specialists responsible for or involved in planning, performing and/or reviewing safety assessments of shallow ground radioactive waste repositories. It discusses approaches that are applicable for safety analysis of a shallow ground repository. The methodologies, analysis techniques and models described are pertinent to the task of predicting the long-term performance of a shallow ground disposal system. They may be used during the processes of selection, confirmation and licensing of new sites and disposal systems or to evaluate the long-term consequences in the post-sealing phase of existing operating or inactive sites. The analysis may point out need for remedial action, or provide information to be used in deciding on the duration of surveillance. Safety analysis both general in nature and specific to a certain repository, site or design concept, are discussed, with emphasis on deterministic and probabilistic studies

  14. A risk-informed perspective on deterministic safety analysis of nuclear power plants

    International Nuclear Information System (INIS)

    Wan, P.T.

    2009-01-01

    In this work, the deterministic safety analysis (DSA) approach to nuclear safety is examined from a risk-informed perspective. One objective of safety analysis of a nuclear power plant is to demonstrate via analysis that the risks to the public from events or accidents that are within the design basis of the power plant are within acceptable levels with a high degree of assurance. This nuclear safety analysis objective can be translated into two requirements on the risk estimates of design basis events or accidents: the nominal risk estimate to the public must be shown to be within acceptable levels, and the uncertainty in the risk estimates must be shown to be small on an absolute or relative basis. The DSA approach combined with the defense-in-depth (DID) principle is a simplified safety analysis approach that attempts to achieve the above safety analysis objective in the face of potentially large uncertainties in the risk estimates of a nuclear power plant by treating the various uncertainty contributors using a stylized conservative binary (yes-no) approach, and applying multiple overlapping physical barriers and defense levels to protect against the release of radioactivity from the reactor. It is shown that by focusing on the consequence aspect of risk, the previous two nuclear safety analysis requirements on risk can be satisfied with the DSA-DID approach to nuclear safety. It is also shown the use of multiple overlapping physical barriers and defense levels in the traditional DSA-DID approach to nuclear safety is risk-informed in the sense that it provides a consistently high level of confidence in the validity of the safety analysis results for various design basis events or accidents with a wide range of frequency of occurrence. It is hoped that by providing a linkage between the consequence analysis approach in DSA with a risk-informed perspective, greater understanding of the limitation and capability of the DSA approach is obtained. (author)

  15. Safety of GM crops: compositional analysis.

    Science.gov (United States)

    Brune, Philip D; Culler, Angela Hendrickson; Ridley, William P; Walker, Kate

    2013-09-04

    The compositional analysis of genetically modified (GM) crops has continued to be an important part of the overall evaluation in the safety assessment program for these materials. The variety and complexity of genetically engineered traits and modes of action that will be used in GM crops in the near future, as well as our expanded knowledge of compositional variability and factors that can affect composition, raise questions about compositional analysis and how it should be applied to evaluate the safety of traits. The International Life Sciences Institute (ILSI), a nonprofit foundation whose mission is to provide science that improves public health and well-being by fostering collaboration among experts from academia, government, and industry, convened a workshop in September 2012 to examine these and related questions, and a series of papers has been assembled to describe the outcomes of that meeting.

  16. Generation IV reactors and the ASTRID prototype: lessons from the Fukushima accident

    International Nuclear Information System (INIS)

    Gauche, F.

    2012-01-01

    In France, the ASTRID prototype is an industrial demonstrator of a sodium-cooled fast neutron reactor (SFR), fulfilling the criteria for Generation IV reactors. ASTRID will meet safety requirements as stringent as for third generation reactors, and it takes into account lessons from the Fukushima accident. The objectives are to reinforce the robustness of the safety demonstration for all safety functions. ASTRID will feature an innovative core with a negative sodium void coefficient, it will take advantage of the large thermal inertia of SFR for decay heat removal, and will provide for a design either eliminating the sodium-water reaction, or guaranteeing no consequences for safety in case such reaction would take place. (author)

  17. Multi-dimensional Code Development for Safety Analysis of LMR

    International Nuclear Information System (INIS)

    Ha, K. S.; Jeong, H. Y.; Kwon, Y. M.; Lee, Y. B.

    2006-08-01

    A liquid metal reactor loaded a metallic fuel has the inherent safety mechanism due to the several negative reactivity feedback. Although this feature demonstrated through experiments in the EBR-II, any of the computer programs until now did not exactly analyze it because of the complexity of the reactivity feedback mechanism. A multi-dimensional detail program was developed through the International Nuclear Energy Research Initiative(INERI) from 2003 to 2005. This report includes the numerical coupling the multi-dimensional program and SSC-K code which is used to the safety analysis of liquid metal reactors in KAERI. The coupled code has been proved by comparing the analysis results using the code with the results using SAS-SASSYS code of ANL for the UTOP, ULOF, and ULOHS applied to the safety analysis for KALIMER-150

  18. Correlation factors for C-14, Cl-36, Ni-59, Ni-63, Mo-93, Tc-99, I-129 and Cs-135. In operational waste for SFR 1

    Energy Technology Data Exchange (ETDEWEB)

    Lindgren, Maria; Pettersson, Michael; Wiborgh, Marie (Kemakta AB, Stockholm (SE))

    2007-01-15

    The use of correlation factors is one possibility to estimate a reference radionuclide inventory for the purpose of safety assessment studies, when waste-type specific information is not at hand. The correlation factor approach requires that there is a correlation between the activity content of the difficult-to-measure nuclide and the key nuclide. A safety assessment of a future deep repository for low and intermediate level waste (SFL 3-5) was conducted during the second half of the 1990's. Within that project, correlation factors for estimating the inventory of radionuclides which are difficult to measure experimentally were developed. These factors have also partly been used to estimate a reference inventory for SFR 1. Based on a literature survey and recent reports published by SKB, it is concluded that new information is available making an update of some correlation factors relevant. For these radionuclides, reported data from estimations and measurements of radionuclide content and correlation factors in different types of radioactive wastes are summarised. The data is evaluated and updated correlation factors representative for fresh waste for eight radionuclides (C-14, Cl-36, Ni-59, Ni-63, Mo-93, Tc-99, I-129 and Cs-135) are suggested. New information from measurements in Swedish NPP has made it possible to define factors that are BWR and PWR specific for C-14, Ni-59 and Ni-63. The uncertainties in suggested data and the applicability of the correlation factors for estimates of the radionuclide content in operational waste are commented upon

  19. Job safety and awareness analysis of safety implementation among electrical workers in airport service company

    Directory of Open Access Journals (Sweden)

    Putra Perdana Suteja

    2018-01-01

    Full Text Available Electrical is a fundamental process in the company that has high risk and responsibility especially in public service company such as an airport. Hence, the company that operates activities in the airport has to identify and control the safety activities of workers. On the safety implementation, the lack of workers’ awareness is fundamental aspects to the safety failure. Therefore, this study aimed to analyse the safety awareness and identify risk in the electrical workplace. Safety awareness questionnaires are distributed to ten workers in order to analyse their awareness. Job safety analysis method used to identify the risk in the electrical workplace. The preliminary study stated that workers were not aware of personal protective equipment usage so that the awareness and behavioural need to be analysed. The result is the hazard was found such as electrical shock and noise for various intensity in the workplace. While electrical workers were aware of safety implementation but less of safety behaviour. Furthermore, the recommendation can be implemented are the implementation of behaviour-based safety (BBS, 5S implementation and accident report list.

  20. Analysis Method of Common Cause Failure on Non-safety Digital Control System

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yun Goo; Oh, Eun Gse [KHNP, Daejeon (Korea, Republic of)

    2014-08-15

    The effects of common cause failure on safety digital instrumentation and control system had been considered in defense in depth analysis with safety analysis method. However, the effects of common cause failure on non-safety digital instrumentation and control system also should be evaluated. The common cause failure can be included in credible failure on the non-safety system. In the I and C architecture of nuclear power plant, many design feature has been applied for the functional integrity of control system. One of that is segmentation. Segmentation defenses the propagation of faults in the I and C architecture. Some of effects from common cause failure also can be limited by segmentation. Therefore, in this paper there are two type of failure mode, one is failures in one control group which is segmented, and the other is failures in multiple control group because that the segmentation cannot defense all effects from common cause failure. For each type, the worst failure scenario is needed to be determined, so the analysis method has been proposed in this paper. The evaluation can be qualitative when there is sufficient justification that the effects are bounded in previous safety analysis. When it is not bounded in previous safety analysis, additional analysis should be done with conservative assumptions method of previous safety analysis or best estimation method with realistic assumptions.